ML15068A045

South Texas Project, Units 1 and 2 - Response to Request for Additional Information Regarding License Amendment Request for Emergency Action Level Scheme Change

ML15068A045
Person / Time
Site:	South Texas
Issue date:	02/26/2015
From:	Capristo A South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-15003226, STI: 34060108, TAC MD4195, TAC MD4196
Download: ML15068A045 (464)

Text

{{#Wiki_filter:Nuclear Operating Company South Texas Prolect Electric Generating Station PO. lat 289 Wadsworth. Teas 77483 -vV/-- --February 26, 2015 NOC-AE-15003226 10 CFR 50.90 File No. G25 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 South Texas Project Units I and 2 Docket Nos. STN 50-498, STN 50-499 Response to Request for Additional Information -South Texas Project (STP), Units 1 and 2 License Amendment Request for Emergency Action Level Scheme Change (TACs MD4195 and MF4196)

References:

1. Letter; G. T. Powell to USNRC Document Control Desk; "License Amendment Request for Revision to Unit 1 and Unit 2 Emergency Action Levels;" NOC-AE-14003087; dated May 15, 2014 (ML14164A341)

2. E-mail; Balwant Singal to Lance Sterling; "Request for Additional Information (RAI) -Revised Emergency Action Levels for South Texas Project, Units 1 and 2 (TACs MD4195 and MF4196);" dated December 18, 2014 (ML14352A1 80)3. Letter; A. Capristo to USNRC Document Control Desk; "Response to Request for Additional Information

-South Texas Project (STP), Units 1 and 2 License Amendment Request for Emergency Action Level Scheme Change (TACs MD4195 and MF4196);" NOC-AE-15003214; dated February 11, 2015 By Reference 1, STP Nuclear Operating Company (STPNOC) requested approval of a License Amendment Request for revision to Unit 1 and 2 Emergency Action Levels. By Reference 2, the NRC staff requested additional information (RAI) to complete its review. STPNOC responded to the RAI in Reference 3 with the exception of RAI-04. STPNOC's response to RAI-04 in Reference 2 is provided in Attachment 1 to this letter. Based on a phone call with NRC staff following transmittal of Reference 3, STPNOC is also including in Attachment 1 revised responses to RAI-01, RAI-06 and RAI-12.In Reference 3, STPNOC noted that several discrepancies were found in the calculation that was performed to determine EAL threshold values for Abnormal Rad Levels. In the process of resolving the discrepancy, a review was conducted of calculations that were performed to support the proposed revised EALs. As a result, six calculations were revised; two of the revised calculations affect STPNOC's response to RAI-04 and RAI-10. The revised response to RAI-10 is included in Attachment

1. Revision bars in Attachment 1 indicate changes to the RAI response provided in Reference 3 for RAI-01, RAI-04, RAI-06, RAI-10, and RAI-12.STI: 34060108 AYqo NOC-AE-1 5003226 Page 2 of 3 A clean copy and a redline markup of the STPEGS Emergency Action Level Technical Bases Document is included in Attachments 2 and 3, respectively.

Attachment 4 provides revisions to the STPEGS Emergency Action Level Deviation, Difference and Justification Matrix provided in Reference

1. Attachment 5 contains revisions to two calculations that are being provided as supporting documents.

The No Significant Hazards Consideration determination provided in Reference I is not altered by the additional information provided in this correspondence. There are no commitments in this letter.If there are any questions, please contact Drew Richards at (361) 972-7666 or me at (361) 972-7697.I declare under penalty of perjury that the foregoing is true and correct.Executed on __ -_____- __________ Date Aldo Capristo Executive Vice President Chief Administrative Officer amr

Attachment:

1. Response to Request for Additional Information

-South Texas Project (STP), Units 1 and 2 License Amendment Request for Emergency Action Level Scheme Change 2. STPEGS Emergency Action Level Technical Bases Document -clean copy 3. STPEGS Emergency Action Level Technical Bases Document -redline markup 4. STPEGS Emergency Action Level Deviation, Difference and Justification Matrix -revisions only 5. Supporting documents NOC-AE-15003226 Page 3 of 3 cc: (paper copy)(electronic copy)Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Lisa M. Regner Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (MS 8 G9A)11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. 0. Box 289, Mail Code: MN1 16 Wadsworth, TX 77483 Morgan, Lewis & Bockius LLP Steve Frantz U.S. Nuclear Regulatory Commission Lisa M. Regner NRG South Texas LP John Ragan Chris O'Hara Jim von Suskil CPS Energqy Kevin Polio Cris Eugster L. D. Blaylock Crain Caton & James, P.C.Peter Nemeth City of Austin Cheryl Mele John Wester Texas Dept. of State Health Services Richard A. Ratliff Robert Free Attachment I Response to Request for Additional Information -South Texas Project (STP), Units 1 and 2 License Amendment Request for Emergency Action Level Scheme Change Attachment 1 NOC-AE-15003226 Page 1 of 5 REQUEST FOR ADDITIONAL INFORMATION SOUTH TEXAS PROJECT, UNITS 1 AND 2 LICENSE AMENDMENT REQUEST FOR EMERGENCY ACTION LEVEL SCHEME CHANGE DOCKET NUMBERS 50-498 AND 499 The NRC staff requires the following additional information to complete its review of the request: RAI-01 Because the information in the basis document can affect emergency classification decision making, NEI 99-01, Revision 6, Section 4.6 contains an expectation that the basis document will be evaluated in accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Paragraph 50.54(q). Please explain how this expectation will be clearly identified to ensure appropriate reviews are conducted for any potential changes to the basis document.Note: The NRC staff does understand that appropriate administrative controls are in place to ensure that changes to Abnormal and Emergency Operating Procedures are screened to determine if an evaluation pursuant to 10 CFR 50.54(q) is required. This RAI is intended to ensure similar controls are in place for the STP EAL Basis Document.STPNOC RESPONSE TO RAI-01 Administrative controls that address evaluating changes to the STP EAL Technical Bases in accordance with 1OCFR50.54 (q) already exist in OPGP05-ZV-0010, Emergency Plan Change.A note has been added to the cover of the STP EAL Technical Bases stating the following: NOTE: Changes to this document require a review under 10CFR50.54 (q) as directed by OPGP05-ZV-0010, Emergency Plan Change. Attachment 1 NOC-AE-15003226 Page 2 of 5 RAI-04 The proposed EALs RU1 appears to be a different base-value than the escalation values (RA1, etc.). Please justify further or revise accordingly. If the values are correct, please note the discrepancy in the basis section.STPNOC RESPONSE TO RAI-04 STPNOC maintains that the RU1 and RA1 values for the Unit Vent radiation monitor (RT-8010B) are consistent with the guidelines of NEI 99-01 Rev.6 and the current EAL scheme at STP. STP does not have Main Steam Line radiation monitor (RT-8046 thru 8049) RU1 and RA1 values in the current EAL scheme for comparison. The proposed STP RU1 thru RG1 EALs were compared to another US PWR (Comanche Peak) and were found to be similar. STPNOC has not identified any discrepancies. The STP Radiation Monitoring system is a distributed microprocessor based system consisting of radiation detectors, an associated microprocessor control panel (RM-80), radiation monitoring computers (RM-1 1), and digital display modules (RM-23) for the Process and Effluents Radiation Monitoring System (PERMS) and the Area Radiation Monitoring System (ARMS).According to the STP UFSAR, the Unit Vent radiation monitor (RT-8010B) has a range from 20 pci/sec to 2E+16 pci/sec and the Main Steam Line radiation monitors have a range from 1.4E-02 pci/cc to 1.4E+06 pci/cc.The hand calculated and STAMPEDE (software) calculated values for the RU1, RA1, RS1 and RG1 are contained in Calculation No. STPNOC013-CALC-002 Revision 2, Table 2.1. Table 2.1 appears in section 2.0, Summary of Results and also in Attachment 1 -Hand Calculations. Attachment 1, Summary of Results states the following: "Table 2.1 is displayed again below showing the results from all the calculations. The minor difference is due to STAMPEDE using decay factors over a one hour period after shutdown. This also accounts for the change in the limiting dose being TEDE in the hand calculation and Thyroid CDE in the STAMPEDE calculations. The accuracy of the hand calculation is considered sufficient and recommended for use in Emergency Action Levels." The Unit Vent (1.40E+05 [ici/sec) and the Main Steam Line (5.00E-02 pci/cc) RU1 values are the unadjusted hand calculated values found in Calculation STPNOCO13-CALC-002 Rev. 2 Table 2.1. The Unit Vent RA1 value (1.50E+06 pci/sec) was derived by conservatively rounding the Table 2.1 hand calculated value of 1.57E+06 pci/sec. The Main Steam Line RA1 value (4.00E+00 pci/cc) was derived by conservatively rounding the hand calculated value of 4.03E+00 pci/cc.The relationship of site boundary doses between the Alert and the Unusual Event is approximately a factor of fifty. The gaps between the Unusual Event and the Alert are expected to be different from the gaps between the Alert and higher classifications based on the underlying assumptions in NEI 99-01.Radon response at STP is expected to be no more than 1% of the UE level based on operating experience and engineering judgment. Attachment 1 NOC-AE-15003226 Page 3 of 5 RAI-06 The proposed EAL RA3.2 includes a number of plant areas for all operating modes. Please verify the plant areas identified for EAL RA3.2 reflect only those areas required for normal plant operations, cooldown, or shutdown, and that access to these areas is required, i.e., cannot be operated remotely. Please provide evidence of this verification, or revise as necessary to support accurate and timely assessment. In addition, consider adding operating mode specificity to the listed areas if applicable. STPNOC RESPONSE TO RAI-06 STPNOC has revised EAL RA3.2 to include a listing of plant areas that require access for normal plant operations, cooldown, or shutdown and components in these areas cannot be remotely operated. Additionally, modes of applicability have been included for each area as follows: (2) An UNPLANNED event results in radiation levels that prohibit or impede access to ANY of the areas listed in Table H3/R2.TABLEEH3/R2: "Pant Areas Requiring Access RCB RHR Heat Exchanger Rooms 0 o MAB 51 ft Room 335 EAB Roof, MCC 1G8, 4.16KV Switchgear Rooms u E 0 0 "n EAB 4.16KV Switchgear Rooms EAL Selection Basis The areas listed in EAL-2 apply to areas that contain equipment necessary for plant operations, cooldown, or shutdown. Assuming all plant equipment is operating as designed, Normal operations and safe shutdown equipment operation is capable from the Main Control Room (MCR). The plant is able to transition into a hot shutdown from the MCR, therefore H3/R2 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary following entry into hot shutdown (establish Residual Heat Removal shutdown cooling, disable operation of charging and ECCS equipment, and limit dilution pathways) and subsequent entry into cold shutdown (disable operation of ECCS equipment). After achieving cold shutdown it is assumed that the plant will be maintained in a cold shutdown condition. Attachment 1 NOC-AE-15003226 Page 4 of 5 RAI-10 The proposed Loss of RCS Barrier due to Category 3, RCS Activity / Containment Radiation, Threshold A, contains a plant-specific basis discussion where temperature induced currents as the result of an RCS leak would preclude the use of containment radiation monitors (RT-8050 and RT-8051) for approximately 40 minutes, and a secondary system break would preclude the use of the containment radiation monitors for 90 minutes.a. Please add this information to the table as the table is the decision-maker tool used for EAL determination, or justify how this information will consistently be used by EAL decision-makers.

b. Please explain why these limitations were not included for Containment Barrier Potential Loss threshold 3.A.1, or revise accordingly.

c. Please explain why these limitations were not included for Fuel Clad Barrier Loss threshold 3.A.1, or revise accordingly.

STPNOC RESPONSE TO RAI-10 STPNOC has removed containment radiation monitors (RT-8050 and RT-8051) from the RCS Barrier Category 3, RCS Activity/Containment Radiation table due to a reduction in the setpoint value based on calculation STPNOC013-CALC-004, Revision 2. Calculation STPNOC013-CALC-004 was revised in February 2015 and would have lowered the RT-8050 and RT-8051 setpoint from 450 mR/hr to approximately 140 mR/hr above background. STPNOC believes that the proximity of the new setpoint to the background level and the effect of TIC precludes the use of these radiation monitors as reliable indications of an RCS breach.STPNOC does not have other Reg. Guide 1.97 radiation monitors in the containment that can fulfill the function of RT-8050 and RT-8051.STPNOC has revised the bases for the Containment Barrier Potential Loss threshold 3.A. 1 and the Fuel Clad Barrier Loss threshold 3.A.1 to state that temperature induced current (TIC) is not a limitation for these events.Temperature induced current (TIC) limitations are not applicable to the Containment Barrier Potential Loss threshold 3.A.1 (Fuel Clad Barrier Loss threshold 3.A. 1) because the expected radiation dose for this event overwhelms the TIC effect. This is discussed in the 1OCFR50.59 evaluation 04-8245-60 associated with DCP 04-8245-33. Attachment 1 NOC-AE-15003226 Page 5 of 5 RAI-12 The proposed EAL HA5 appears to cover a wide range of rooms or areas during all modes of operation.

a. Please verify the plant areas identified for EAL RA3.2 reflect only those areas required for normal plant operations, cooldown, or shutdown, and that access to these areas is required, i.e., cannot be operated remotely.

Please provide evidence of this verification, or revise as necessary to support accurate and timely assessment. In addition, consider adding operating mode specificity to the listed areas if applicable.

b. For EAL HA5, please provide justification for the omission of the control room as a plant area where access is needed to support normal plant operations, cooldown, or shutdown.STPNOC RESPONSE TO RAI-12 STPNOC has revised EAL HA5.1a to include the control room as a plant area where access is needed to support normal plant operations, cooldown, or shutdown.1 a. Release of a toxic, corrosive, asphyxiant or flammable gas into the Control Room or any of the plant rooms or areas listed in Table H3/R2: AND b. Entry into the room or area is prohibited or impeded._ TABLE H3/R2: Plant Areas Re6uiring Access RCB RHR Heat Exchanger Rooms 0 o 0 MAB 51 ft Room 335 EAB Roof, MCC 1G8, 4.16KV Switchgear Rooms Lu 0 o Ln EAB 4.16KV Switchgear Rooms EAL Selection Basis: The areas listed in EAL-1 apply to areas that contain equipment necessary for plant operations, cooldown, or shutdown.

Assuming all plant equipment is operating as designed, Normal operations and safe shutdown equipment operation is capable from the Main Control Room (MCR). The plant is able to transition into a hot shutdown from the MCR, therefore H3/R2 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary following entry into hot shutdown (establish Residual Heat Removal shutdown cooling, disable operation of charging and ECCS equipment, and limit dilution pathways) and subsequent entry into cold shutdown (disable operation of ECCS equipment). After achieving cold shutdown it is assumed that the plant will be maintained in a cold shutdown condition. Attachment 2 STPEGS Emergency Action Level Technical Bases Document -clean copy STPEGS Emergency Action Level Technical Bases Document Rev. 0 NEI 99-01 Rev. 6 Implementation February 2015 NOTE: Changes to this document require a review under 10CFR50.54 (q) as directed by OPGP05-ZV-O010, Emergency Plan Change. TABLE OF CONTENTS 1 DEVELOPMENT OF EMERGENCY ACTION LEVELS ........ Error! Bookmark not defined.1.1 REGULATORY BACKGROUND ................................................................................................ 1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) .......................................... 1 1.3 N R C O R D E R E A 05 1 ................................................................................................................... 2 2 KEY TERMINOLOGY ......................................................................................................... 3 2.1 EMERGENCY CLASSIFICATION LEVEL (ECL) ...................................................................... 3 2.2 IN ITIA TIN G CO ND IT IO N (IC) ................................................................................................... 5 2.3 EM ERGEN CY A CTION LEVEL (EAL) ........................................................................................... 5 2.4 FISSION PRODUCT BARRIER THRESHOLD ............................................................................... 5 3 DESIGN OF THE STPEGS EMERGENCY CLASSIFICATION SCHEME ........................... 7 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS) ............................... 7 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS ................. 10 3.3 STPEGS DESIGN CONSIDERATIONS ..................................................................................... 10 3.4 ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION ............................... 11 3.5 IC AND EAL M ODE APPLICABILITY ..................................................................................... 11 4 STPEGS SCHEME DEVELOPMENT ............................................................................ 13 4.1 GENERAL DEVELOPMENT PROCESS ................................................................................... 13 4.2 CRITICA L CH A RA CTERISTICS ............................................................................................... 13 4.3 INSTRUMENTATION USED FOR EALS ................................................................................ 13

4.4 REFERENCES

TO STPEGS AOPS AND EOPS ........................................................................ 14 5 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ........................................ 14 5.1 GENERA L CON SIDERA TION S ............................................................................................... 14 5.2 CLASSIFICATION M ETHODOLOGY ....................................................................................... 15 5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS ........................................ 15 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION ............................. 16 5.5 CLASSIFICATION OF IMMINENT CONDITIONS ................................................................. 16 5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING ............. 16 5.7 CLASSIFICATION OF SHORT-LIVED EVENTS .................................................................... 17 5.8 CLASSIFICATION OF TRANSIENT CONDITIONS ................................................................ 17 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION ............... 18 5.10 RETRACTION OF AN EMERGENCY DECLARATION ....................................................... 18 6 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/EALS ......................... 19 7 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS ...................... 40 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS .............. 63 9 FISSION PRODUCT BARRIER ICS/EALS .................................................................. 66 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ........... 85 11 SYSTEM MALFUNCTION ICS/EALS ........................................................................... 111 APPENDIX A -ACRONYMS AND ABBREVIATIONS ....................................................... 140 APPENDIX B -DEFINITIONS .......................................................................................... 142 THIS PAGE IS LEFT INTENTIONALLY BLANK 1 DEVELOPMENT OF EMERGENCY ACTION LEVELS 1.1 REGULATORY BACKGROUND Title 10, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC)regulations that apply to nuclear power facilities. Several of these regulations govern various aspects of an emergency classification scheme. A review of the relevant sections listed below will aid the reader in understanding the key terminology provided in Section 3.0 of this document.1 10 CFR § 50.47(a)(1)(i) 1 10 CFR § 50.47(b)(4) 1 10 CFR § 50.54(q)1 10 CFR § 50.72(a)1 10 CFR § 50, Appendix E, IV.B, Assessment Actions* 10 CFR § 50, Appendix E, IV.C, Activation of Emergency Organization The above regulations are supplemented by various regulatory guidance documents. Three documents of particular relevance to NEI 99-01 are:* NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants]" NUREG-1022, Event Reporting Guidelines 10 CFR § 50.72 and § 50.73 Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactor 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)South Texas Project Electrical Generating Station (STP or STPEGS) is locating an ISFSI approximately 450 feet west of the Unit 2 Reactor Building. The STP ISFSI will be within the site Protected Area and is scheduled to be operational in 2016.Selected guidance in NEI 99-01 is applicable to the STPEGS emergency plan to fulfill the requirements of 10 CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CFR § 50 and the guidance in NUREG 0654/FEMA-REP-1. The initiating conditions germane to a 10 CFR § 72.32 emergency plan (as described in NUREG-1567) are subsumed within the classification scheme for a 10 CFR § 50.47 emergency plan.The STPEGS ICs and EALs for an ISFSI are presented in Section 8, ISFSI ICs/EALs. IC E-HUI covers the spectrum of credible natural and man-made events included within the scope of the STPEGS ISFSI design. In addition, appropriate aspects of IC HU1 and IC HAl address a HOSTILE ACTION directed against the STPEGS ISFSI.II P ag e The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG- 1140, A Regulatory Analysis on Eme-ge/ic)' Preparedness for Fuel Cycle and Other Radioactive MIaterial Licensees. NUREG- 1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 rem Effective Dose Equivalent. Regarding the above information, the expectations for an offsite response to an ALERT classified tunder a 10 CFR § 72.32 emergency plan are generally consistent with those for an UNUSUAL EVENT in a 10 CFR § 50.47 emergency plan (e.g., to provide assistance if requested). Also, the STPEGS Emergency Response Organization (ERO) required for a 10 CFR § 72.32 emergency plan is different than that prescribed for a 10 CFR § 50.47 emergency plan (e.g., no emergency technical support function). 1.3 NRC ORDER EA-12-051 The Fukushima Daiichi accident of March 11, 2012, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity, and ultimately led to core damage in three reactors. While the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to preclude fuel damage from the loss of cooling.Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii). Among them was to provide each spent fuel pool with reliable level instrumentation to significantly enhance the ability of key decision-makers to allocate resources effectively following a beyond design basis event. To this end, the NRC issued Order EA-12-051. Issuance of Order to Modio, Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all US nuclear plants with an operating license, construction permit, or combined construction and operating license.NRC Order EA-12-051 states, in part, "All licensees ... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel: (1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." To this end, all licensees must provide:* A primary and back-up level instrument that will monitor water level from the nonnal level to the top of the used fuel rack in the pool;* A display in an area accessible following a severe event; and* Independent electrical power to each instrument channel and provide an alternate remote power connection capability. NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", provides guidance for complying with NRC Order EA-12-051. 2 P age This document includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA- 12-05 1. These EALs are included within existing IC RA2, and new ICs RS2 and RG2. These EALs will be implemented when the enhanced spent fuel pool level instrumentation is available for use.2 KEY TERMINOLOGY USED There are several key terms that appear throughout the EAL methodology. These terms are introduced in this section to support understanding of subsequent material. As an aid to the reader, the following table is provided as an overview to illustrate the relationship of the terms to each other.EMERGENCY CLASSIFICATION LEVEL UNUSUAL EVENT ALERT SAE GE 4, 40 +Initiating Condition Initiating Condition Initiating Condition Initiating Condition 4, + 4' +" Emergency Action Emergency Action Emergency Action Emergency Action Level (1) Level (1) Level (1) Level (1)" Operating Mode -Operating Mode

• Operating Mode
• Operating Mode Applicability Applicability Applicability Applicability" Notes -Notes -Notes
• Notes" Basis
• Basis -Basis
• Basis (1) -When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition.

This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basis information. In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL.2.1 EMERGENCY CLASSIFICATION LEVEL (ECL)One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The EMERGENCY CLASSIFICATION LEVELS, in ascending order of severity, are:* UNUSUAL EVENT (UE)* ALERT* SITE AREA EMERGENCY (SAE)* GENERAL EMERGENCY (GE)3 P a o e

2.1.1 UNUSUAL

EVENT (UE)Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.Purpose: The purpose of this classification is to assure that the first step in future response has been carried out, to bring the operations staff to a state of readiness, and to provide systematic handling of unusual event information and decision-making.

2.1.2 ALERT

Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening, risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels.Purpose: The purpose of this classification is to assure that emergency personnel are readily available to respond if the situation becomes more serious or to perform confirmatory radiation monitoring if required, and provide offsite authorities current information on plant status and parameters. 2.1.3 SITE AREA EMERGENCY (SAE)Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.Purpose: The purpose of the SITE AREA EMERGENCY declaration is to assure that emergency response centers are staffed, to assure that monitoring teams are dispatched, to assure that personnel required for evacuation of near-site areas are at duty stations if the situation becomes more serious, to provide consultation with offsite authorities, and to provide updates to the public through government authorities.

2.1.4 GENERAL

EMERGENCY (GE)Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.Purpose: The purpose of the GENERAL EMERGENCY declaration is to initiate predetermined protective actions for the public, to provide continuous assessment of information from the licensee and offsite organizational measurements, to initiate additional measures as indicated by actual or potential releases, to provide consultation with offsite authorities, and to provide updates for the public through government authorities. 4 1P a oe

2.2 INITIATING

CONDITION (IC)An event or condition that aligns with the definition of one of the four EMERGENCY CLASSIFICATION LEVELS by virtue of the potential or actual effects or consequences. Discussion: An IC describes an event or condition, the severity or consequences of which meets the definition of an emergency classification level. An IC can be expressed as a continuous, measurable parameter (e.g., RCS leakage), an event (e.g., an earthquake) or the status of one or more fission product barriers (e.g., loss of the RCS barrier).Appendix I of NUREG-0654 does not contain example Emergency Action Levels (EALs) for each ECL, but rather Initiating Conditions (i.e., plant conditions that indicate that a radiological emergency, or events that could lead to a radiological emergency, has occurred). NUREG-0654 states that the Initiating Conditions forn the basis for establishment by a licensee of the specific plant instrumentation readings (as applicable) which, if exceeded, would initiate the emergency classification. Thus, it is the specific instrument readings that would be the EALs.Considerations for the assignment of a particular INITIATING CONDITION to an EMERGENCY CLASSIFICATION LEVEL are discussed in Section 3.2.2.1 EMERGENCY ACTION LEVEL (EAL)A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level.Discussion: EAL statements may utilize a variety of criteria including instrument readings and status indications; observable events; results of calculations and analyses; entry into particular procedures; and the occurrence of natural phenomena.

2.2.2 FISSION

PRODUCT BARRIER THRESHOLD A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.Discussion: Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are:* Fuel Clad* Reactor Coolant System (RCS)* Containment Upon determination that one or more fission product barrier thresholds have been exceeded, the combination of barrier loss and/or potential loss thresholds is compared to the fission product barrier IC/EAL criteria to determine the appropriate ECL.5 1 P age In some accident sequences, the ICs and EALs presented in the Abnormal Radiation Levels/ Radiological Effluent (R) Recognition Category will be exceeded at the same time, or shortly after, the loss of one or more fission product barriers. This redundancy is intentional as the former ICs address radioactivity releases that result in certain offsite doses from whatever cause, including events that might not be fully encompassed by fission product barriers (e.g., spent fuel pool accidents, design containment leakage following a LOCA, etc.).6 1P a g e 3 DESIGN OF THE STPEGS EMERGENCY CLASSIFICATION SCHEME 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS)An effective emergency classification scheme must incorporate a realistic and accurate assessment of risk, both to plant workers and the public. There are obvious health and safety risks in underestimating the potential or actual threat from an event or condition; however, there are also risks in overestimating the threat as well (e.g., harm that may occur during an evacuation). The emergency classification scheme attempts to strike an appropriate balance between reasonably anticipated event or condition consequences, potential accident trajectories, and risk avoidance or minimization. There are a range of "non-emergency events" reported to the US Nuclear Regulatory Commission (NRC) staff in accordance with the requirements of 10 CFR § 50.72. Guidance concerning these reporting requirements, and example events, are provided in NUREG-1022. Certain events reportable under the provisions of 10 CFR §50.72 may also require the declaration of an emergency. In order to align each Initiating Conditions (IC) with the appropriate ECL, it was necessary to determine the attributes of each ECL. The goal of this process is to answer the question, "What events or conditions should be placed tinder each ECL?" The following sources provided information and context for the development of ECL attributes.

• Assessments of the effects and consequences of different types of events and conditions
• STPEGS abnormal and emergency operating procedure setpoints and transition criteria* STPEGS Technical Specification limits and controls* STPEGS Offsite Dose Calculation Manual (ODCM) radiological release limits* Review of selected STPEGS Updated Final Safety Analysis Report (UFSAR) accident analyses* Enviromnental Protection Agency (EPA) Protective Action Guidelines (PAGs)* NUREG 0654, Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants* Industry Operating Experience
• Input from subject matter experts at STPEGS The following ECL attributes were created to aid in the development of ICs and Emergency Action Levels (EALs). The attributes may be useful in briefing and training settings (e.g., helping an Emergency Director understand why a particular condition is classified as an ALERT).7 1 a oe The attributes of each ECL are presented below.3.1.1 UNUSUAL EVENT (UE)An UNUSUAL EVENT, as defined in section 2.1.1, includes but is not limited to an event or condition that involves: (A) A precursor to a more significant event or condition.(B) A minor loss of control of radioactive materials or the ability to control radiation levels within the plant.(C) A consequence otherwise significant enough to warrant notification to local, State and Federal authorities.

3.1.2 ALERT

An ALERT, as defined in section 2.1.2, includes but is not limited to an event or condition that involves: (A) A loss or potential loss of either the fuel clad or Reactor Coolant System (RCS) fission product barrier.(B) An event or condition that significantly reduces the margin to a loss or potential loss of the fuel clad or RCS fission product barrier.(C) A significant loss of control of radioactive materials resulting in an inability to control radiation levels within the plant, or a release of radioactive materials to the environment that could result in doses greater than 1% of an EPA PAG at or beyond the site boundary.(D) A HOSTILE ACTION occurring within the OWNER CONTROLLED AREA.3.1.3 SITE AREA EMERGENCY (SAE)A SITE AREA EMERGENCY, as defined in section 2.1.3, includes but is not limited to an event or condition that involves: (A) A loss or potential loss of any two fission product barriers -fuel clad, RCS and/or containment.(B) A precursor event or condition that may lead to the loss or potential loss of multiple fission product barriers within a relatively short period of time. Precursor events and conditions of this type include those that challenge the monitoring and/or control of multiple safety systems.(C) A release of radioactive materials to the environment that could result in doses greater than 10% of an EPA PAG at or beyond the site boundary.(D) A HOSTILE ACTION occurring within the plant PROTECTED AREA.8 I P a g e

3.1.4 GENERAL

EMERGENCY (GE)A GENERAL EMERGENCY, as defined in section 2.1.4, includes but is not limited to an event or condition that involves: (A) Loss of any two fission product barriers AND loss or potential loss of the third barrier -fuel clad, RCS and/or containment.(B) A precursor event or condition that, Unmitigated, may lead to a loss of all three fission product barriers. Precursor events and conditions of this type include those that lead directly to core damage and loss of containment integrity.(C) A release of radioactive materials to the environment that could result in doses greater than an EPA PAG at or beyond the site boundary.(D) A HOSTILE ACTION resulting in the loss of key safety functions (reactivity control, core cooling/RPV water level or RCS heat removal) or damage to spent fuel.3.1.5 Risk-Informed Insights Emergency preparedness is a defense-in-depth measure that is independent of the assessed risk from any particular accident sequence; however, the development of an effective emergency classification scheme can benefit from a review of risk-based assessment results. To that end, the development and assignment of certain ICs and EALs also considered insights from several site-specific probabilistic safety assessments (PSA -also known as probabilistic risk assessment, PRA). Some generic insights friom this review included: 1. Accident sequences involving a prolonged loss of all AC power are significant contributors to core damage frequency. For this reason, a loss of all AC power for greater than 15 minutes, with the plant at or above Hot Shutdown, was assigned an ECL of SITE AREA EMERGENCY. Precursor events to a loss of all AC power were also included as an UNUSUAL EVENT and an ALERT.A station blackout coping analyses performed in response to 10 CFR § 50.63 and Regulatory Guide 1.155, Station Blackout, may be used to determine a time-based criterion to demarcate between a SITE AREA EMERGENCY and a GENERAL EMERGENCY. The time dimension is critical to a properly anticipatory emergency declaration since the goal is to maximize the time available for State and local officials to develop and implement offsite protective actions. STP is an Alternate AC plant and a Station Blackout battery copying analysis is not required. Nonetheless, a 125 VDC Battery Four Hour Coping Analysis was conducted and provides a basis for the time-based escalation path from a SITE AREA EMERGENCY to a GENERAL EMERGENCY.

2. For severe core damage events, uncertainties exist in phenomena important to accident progressions leading to containment failure. Because of these uncertainties, predicting the status of containment integrity may be difficult under severe accident conditions.

This is why maintaining containment integrity alone following sequences leading to severe core damage is an insufficient basis for not escalating to a GENERAL EMERGENCY.

3. PSAs indicated that leading contributors to latent fatalities were sequences involving a containment bypass, a large Loss of Coolant Accident (LOCA) with early containment failure, a Station Blackout lasting longer than four hours, and a reactor coolant pump seal failure. The generic EAL methodology needs to be sufficiently rigorous to address these sequences in a timely fashion.91 P age

3.2 TYPES

OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS The STPEGS methodology makes use of symptom-based, barrier-based and event-based ICs and EALs. Each type is discussed below.Symptom-based ICs and EALs are parameters or conditions that are measurable over some range using plant instrumentation (e.g., core temperature, reactor coolant level, radiological effluent, etc.). When one or more of these parameters or conditions are off-normal, reactor operators will implement procedures to identify the probable cause(s) and take corrective action.Fission product barrier-based ICs and EALs are the subset of symptom-based EALs that refer specifically to the level of challenge to the principal barriers against the release of radioactive material from the reactor core to the environment. These barriers are the fuel cladding, the reactor coolant system pressure boundary, and the containment. The barrier-based ICs and EALs consider the level of challenge to each individual barrier -potentially lost and lost -and the total number of barriers under challenge. Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. These include the failure of an automatic reactor scram/trip to shut down the reactor, natural phenomena (e.g., an earthquake), or man-made hazards such as a toxic gas release.3.3 STPEGS DESIGN CONSIDERATIONS The South Texas Project Electrical Generating Station (STPEGS) is composed of two units, each having an identical pressurized water reactor (PWR) Nuclear Steam Supply System (NSSS) and turbine generator (TG).The NSSS is a Westinghouse Electric Corporation four-loop PWR. High-pressure light water serves as the coolant, neutron moderator, reflector, and solvent for the neutron absorber. The Reactor Coolant System (RCS), comprised of four parallel loops (each with a RCP and a steam generator [SG]), is used to transfer the heat generated in the core to the SGs using RCPs to circulate the water. RCS pressure is maintained by means of a pressurizer attached to the hot leg of one of the loops. The RCS is designed to circulate borated demineralized water at temperatures, pressures and flow rates consistent with the design thermal and hydraulic performance of the NSSS.The Reactor Coolant Pressure Boundary Leak Detection System consists of temperature, level, humidity, and radioactivity sensors with associated instrumentation and alarms. Small leaks are detected by temperature and level changes of systems, increasing sump levels, and humidity and radioactivity concentration changes inside the Containment. Large leaks are detected by changes in reactor coolant inventory, changes in flow rates in process lines and changes in sump level.Emergency Core Cooling System consists of three independent trains, each one capable of providing 100 percent of the required flow to the core in the unlikely event of a LOCA. Each train consists of one high-head safety injection pump and one low-head safety injection pump. Heat is removed from the system during recirculation by the residual heat removal heat exchanger (low-head pump only). The piping and valving associated with each of the three subsystems are identical. In the event of a steam pipe rupture, the ECCS provides adequate shutdown capability. 10 I P a g e The Reactor Containment is a post-tensioned concrete cylinder with a steel liner plate, hemispherical top, and flat bottom. This structure provides a virtually leaktight barrier to prevent escape of fission products to the environment in the unlikely event of a loss of coolant accident (LOCA).3.4 ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION The scheme's generic information is organized by Recognition Category in the following order.* R- Abnormal Radiation Levels / Radiological Effluent -Section 6* C -Cold Shutdown / Refueling System Malfunction -Section 7* E -Independent Spent Fuel Storage Installation (ISFSI) -Section 8* F -Fission Product Barrier -Section 9* H -Hazards and Other Conditions Affecting Plant Safety -Section 10* S -System Malfunction -Section 11 Each Recognition Category section contains a matrix showing the ICs and their associated EMERGENCY CLASSIFICATION LEVELS. The following information and guidance is provided for each IC:* ECL -the assigned emergency classification level for the IC.* Initiating Condition -provides a summary description of the emergency event or condition.

• Operating Mode Applicability

-Lists the modes during which the IC and associated EAL(s) are applicable (i.e., are to be used to classify events or conditions).

• Emergency Action Level(s) -Provides indications that are considered to meet the intent of the IC.For Recognition Category F, the fission product barrier thresholds are presented in tables and arranged by fission product barrier and the degree of barrier challenge (i.e., potential loss or loss). This presentation method shows the synergism among the thresholds, and supports accurate assessments.

Basis -Provides background information that explains the intent and application of the IC and EALs. In some cases, the basis also includes relevant source information and references. 3.5 IC AND EAL MODE APPLICABILITY The STPEGS emergency classification scheme was developed recognizing that the applicability of ICs and EALs will vary with plant mode. For example, some symptom-based ICs and EALs can be assessed only during the power operations, startup, or hot standby/shutdown modes of operation when all fission product barriers are in place, and plant instrumentation and safety systems are fully operational. In the cold shutdown and refueling modes, different symptom-based ICs and EALs will come into play to reflect the opening of systems for routine maintenance, the unavailability of some safety system components and the use of alternate instrumentation. The following table shows which Recognition Categories are applicable in each plant mode. The ICs and EALs for a given Recognition Category are applicable in the indicated modes.I I I P a e MODE OF APPLICABILITY MATRIX Recognition Categorv Mode R C E F H S Power Operations X X X X X Startup X X X X X Hot Standby X X X X X Hot Shutdown X X X X X Cold Shutdown X X X X Refueling X X X X Defueled X X X X STPEGS Operating Modes Mode Description Criteria (Rx Power excludes decay heat)1 Power Operations [ Reactor Power> 5%, Keff> 0.99 T Avg> 350'F 2 Startup [Reactor Power < 5%, Keff> 0.99 T Avg > 350'F[ Hot Standby Reactor Power 0% Keff< 0.99 T Avg> 350'F 4 Hot Shutdown Reactor Power 0% Keff < 0.99 350'F > T Avg > 200'F 5 Cold Shutdown Reactor Power 0% Keff < 0.99 T Avg < 200'F 6 Refueling Reactor Power 0% Keff< 0.95 T Avg < 140°F Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.Defueled All fuel removed from the reactor vessel (i.e., full core offload during refuel or extended outage)12 1 P ag- e 4 STPEGS SCHEME DEVELOPMENT

4.1 GENERAL

DEVELOPMENT PROCESS Tile STPEGS ICs and EALs were developed to be unambiguous and readily assessable because both serve specific purposes. The IC is the fundamental event or condition requiring a declaration. The EAL(s) is the pre-determined threshold that defines when the IC is met. To this end, the STPEGS ICs and EALs were developed with input from key stakeholders such as Operations, Training, Health Physics, and Engineering. STPEGS specific indications, parameters and values were consistent with licensing basis documents, plant procedures, training, calculations, and drawings Useful acronyms and abbreviations associated with the STPEGS emergency classification scheme are presented in Appendix A, Acronyms and Abbreviations. Those specific to STPEGS were included to be consistent with site terminology, site procedure, and training.Many words or terms used in the STPEGS emergency classification scheme have scheme-specific definitions. These words and terms are identified by being set in all capital letters (i.e., ALL CAPS). The definitions are presented in Appendix B, Definitions.

4.2 CRITICAL

CHARACTERISTICS When crafting the scheme, STPEGS ensured that certain critical characteristics were met. These critical characteristics are listed below.The ICs,. EALs, Operating Mode Applicability criteria, Notes and Basis information are consistent with industry guidance; while the actual wording may be different from NEI 99-01 Revision 6, the classification intent is maintained. With respect to Recognition Category F, the STPEGS scheme included a user-aid to facilitate timely and accurate classification of fission product barrier losses and/or potential losses. The user-aid logic is consistent with the classification logic presented in Section 9.* EAL statements use objective criteria and observable values.* ICs, EALs, Operating Mode Applicability and Note statements and formatting consider human factors and are user-friendly.

• The scheme facilitates upgrading and downgrading of the emergency classification where necessary.
• The scheme facilitates classification of multiple concurrent events or conditions.

4.3 INSTRUMENTATION

USED FOR EALS STPEGS incorporated instrumentation that is reliable and routinely maintained in accordance with site programs and procedures. Alarms referenced in EAL statements are those that are the most operationally significant for the described event or condition. EAL setpoints are within the calibrated range of the referenced instrumentation, and consider any automatic instrumentation functions that may impact accurate EAL assessment. In addition, EAL setpoint values do not use terms such as "off-scale low" or "off-scale high" since that type of reading may not be readily differentiated from an instrument failure. If instrumentation failures occur that have EALs associated with them (i.e., process radiation monitors) compensatory means of implementation may be used as described in plant procedures. 13 1 P a g e

4.4 REFERENCES

TO STPEGS AOPS AND EOPS Some of the criteria/values used in several EALs and fission product barrier thresholds were drawn from STPEGS AOPs and EOPs. This approach was intended to maintain good alignment between operational diagnoses and emergency classification assessments. STPEGS verified the appropriate administrative controls are in place to ensure that a subsequent change to anl AOP or EOP is screened to determine if an evaluation pursuant to 10 CFR 50.54(q) is required.5 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS

5.1 GENERAL

CONSIDERATIONS When making anl emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of anl Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basis information. In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL.NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-0 1, Interim Staff Guidance, Emergencv Planning for Nuclear Power Plants.All emergency classification assessments should be based upon VALID indications, reports or conditions. A VALID indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, validation could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. The validation of indications should be completed in a manner that supports timely emergency declaration. For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.). the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that 1) the activity proceeds as planned and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating 14 1P ag e license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 § CFR 50.72.The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.); the EAL and/or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift). While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. This scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL)definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated into the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.5.2 CLASSIFICATION METHODOLOGY To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL(s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, then the IC is considered met and the associated ECL is declared in accordance with plant procedurss. When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock." For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01.

5.3 CLASSIFICATION

OF MULTIPLE EVENTS AND CONDITIONS When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:* If an ALERT EAL and a SITE AREA EMERGENCY EAL are met, whether at one unit or at two different units, a SITE AREA EMERGENCY should be declared.There is no "additive" effect from multiple EALs meeting the same ECL. For example:* If two ALERT EALs are met, whether at one unit or at two different units, an ALERT should be declared.Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events.15 1 P age

5.4 CONSIDERATION

OF MODE CHANGES DURING CLASSIFICATION The mode in effect at the time that all event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher.5.5 CLASSIFICATION OF IMMINENT CONDITIONS Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in thejudgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all EMERGENCY CLASSIFICATION LEVELS, this approach is particularly important at the higher EMERGENCY CLASSIFICATION LEVELS since it provides additional time for implementation of protective measures.5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated. The following approach to downgrading or terminating an ECL is recommended. ECL Action When Condition No Longer Exists UNUSUAL EVENT Terminate the emergency in accordance with plant procedures. ALERT Downgrade or terminate the emergency in accordance with plant procedures. SITE AREA EMERGENCY with no long- Downgrade or terminate the emergency in term plant damage accordance with plant procedures. SITE AREA EMERGENCY with long- Terminate the emergency and enter term plant damage recovery in accordance with plant procedures. GENERAL EMERGENCY Terminate the emergency and enter recovery in accordance with plant procedures. 16 1 P ag e As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02.5.7 CLASSIFICATION OF SHORT-LIVED EVENTS As discussed in Section 3.2, event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include a failure of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip or an earthquake.

5.8 CLASSIFICATION

OF TRANSIENT CONDITIONS Many of the ICs and/or EALs contained in this document employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions. EAL momentarily met during expected plant response -In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures. EAL momentarily met but the condition is corrected prior to an emergency declaration -If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example.An ATWS occurs and the auxiliary feedwater system fails to automatically start. Steam generator levels rapidly lower and the plant enters an inadequate RCS heat removal condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts the auxiliary feedwater system in accordance with an EOP step and clears the inadequate RCS heat removal condition prior to an emergency declaration, then the classification should be based on the ATWS only.It is important to stress that the 15-minute emergency classification assessment period is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event; emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations where an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions. 17 ..aLe 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements. 5.10 RETRACTION OF AN EMERGENCY DECLARATION Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022. 18 1 P a o e 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS Table R-1: Recognition Category "R" Initiating Condition Matrix UNUSUAL EVENT RUl Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.Op. Modes: All RU2 UNPLANNED loss of water level above irradiated fuel.Op. Modes: All ALERT SITE AREA EMERGENCY GENERAL EMERGENCY RA1 Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrern TEDE or 50 mrem THYROID CDE.Op. Modes: All RA2 Significant lowering of water level above, or damage to, irradiated fuel.Op. Modes: All RA3 Radiation levels that impede access to equipment necessary for nonnal plant operations, cooldown or shutdown.Op. Modes: All RS1 Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrern THYROID CDE.Op. Modes: All RS2 Spent fuel pool level at 40'-4" or lower.Op. Modes: All RG1 Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem THYROID CDE.Op. Modes: All RG2 Spent fuel pool level cannot be restored to at least 40'-4" for 60 minutes or longer.Op. Modes: All 191 Page RU1 ECL: UNUSUAL EVENT Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.Operating Mode Applicability: ALL Emergency Action Levels: (1 or 2 or 3)Notes:* The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.(1) Reading on ANY of the following radiation monitor greater than the values listed in Table RI column"UE" for 60 minutes or longer: Table RI: Effluent Monitors Release Point Monitor GE SAE ALERT UE Unit Vent RT-80 I OB 1.50 E+08 jaCi/sec 1.50 E+07 pCi/sec 1.50 E+06 pCi/sec 1.40 E+05 tCi/sec Main Steam RT-8046 thni 4.00 E+02 ptCi/cm 3 4.00 E+01 pCi/cm 3 4.00 E+00 ptCi/cm 3 5.00 E-02 pCi/cm 3 Lines 8049 (2) Reading on gaseous effluent radiation monitor RT-801 OB greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.(3) Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODCM limits for 60 minutes or longer.Basis: This IC addresses a potential lowering in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.STPEGS incorporated design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.20 1 P a g e Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Classification based onl effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.EAL #1- This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.EAL #2- This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).EAL #3- This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).Escalation of the emergency classification level would be via IC RAI.RUI: EAL-1 Selection Basis The Unit Vent and Main Steam Line monitor readings were included in this EAL because they give instantaneous indications of a monitored gaseous release exceeding twice the ODCM limits. Normal gaseous effluents are due to planned RCB purges and monitored by the Unit Vent. The Main Steam Line monitor readings were included because they correspond to a concentration that would result in a release rate of twice the ODCM limits if there were a release via the Power Operated Relief Valves (PORVs) or Safety Relief Valves. A release from the PORVs or Safety Relief Valves is not a normal effluent pathway but ineets the intent of the EAL.The Unit Vent and Main Steam Line release values are based on Calculation No. STPNOCOI3-CALC-002,. Rev.211 P a,,e RUI: EAL-2. 3 Selection Basis For EAL-2, there are two effluent radiation monitors, RT-8038 (liquid) and RT-800 OB (gaseous), however only RT-8010B was included. The alarm setpoint for the gaseous effluent radiation monitor RT-8010B is set at the ODCM limits. An indication of two times the alarm setpoint (two times the ODCM limit) would allow operators time to secure the release prior to meeting this EAL. The liquid effluent radiation monitor RT-8038 was not included in EAL-2 because the activity in liquid discharges is normally the several orders of magnitude lower than the ODCM limits. In order to alert personnel to significant changes in the liquid effluent activity, the alarm setpoint for RT-8038 is normally set several orders of magnitude below the ODCM limits. Setting the alarm setpoint for RT-8038 at the ODCM limit would remove this capability and violate the intent of the EAL.For EAL-3., sample analysis could be used as a backup for the effluent monitor indications.

REFERENCES:

1. Calculation No: STPNOCO13-CALC-002 Rev. 2, Radiological Release Thresholds for Emergency Action Levels 2. Offsite Dose Calculation Manual (ODCM), Rev. 17, Part B3.0 to B4.9 3. UFSAR, Rev. 14, Section 11.5.2.3.3 and 11.5.2.5.3 (monitor descriptions)

4. UFSAR, Rev. 14, Section 11.5.2.4.4 (liquid waste processing monitor)22 1 P a o e RU2 ECL: UNUSUAL EVENT Initiating Condition:

UNPLANNED loss of water level above irradiated fuel.Operating Mode Applicability: ALL Emergency Action Level: (1) a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

• Visual Observation OR* Annunciator alann on larnpbox 22M02 Window F-5 "SFP WATER LVL HI/LO" OR* Spent fuel in the ICSA AND Annunciator alarm on lampbox 22M02 Window F-6 "SFP Trouble" AND Plant Computer point FCLC1420 "REFLNG CAV LVL IN CNTMT" (ICSA Water Level HI/LO) is in alarm AND b. UNPLANNED rise in area radiation levels on ANY of the following radiation monitors.* RE-8055 (68' RCB) -Mode 5 or 6 only OR" RE-8099 (68' RCB) -Mode 5 or 6 only OR" RE-8090 (68' FHB)Basis: This IC addresses a lowering in water level above irradiated fuel sufficient to cause elevated radiation levels.This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.A water level lowering will be primarily determined by indications from available level instrumentation.

Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations. A significant drop in the water level may also cause a rise in the radiation levels of adjacent areas that can be detected by monitors in those locations. The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may rise due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level.23 1 P a -ge A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC RA2.RU2: EAL-1 Selection Basis Hi/Lo level sensors are located in the Spent Fuel Pool (LSHL 1401) and the RCB, In Containment Storage Area (ICSA) (LSHL 1420). If level in the Spent Fuel Pool rises or lowers by more than 6 inches above or below the normal water level of 66'-6" (UFSAR 9.1.2.1), the "SFP WATER LEVEL H I/LO" lampbox 22M02 window F-5 annunciator alarm is received in the Control Room (0POP09-AN-22M2, Annunciator Lampbox 22M02 Response Instructions). Although the ICSA has a Hi/LO level sensor, there is not an annunciator in the Control Room similar to the one for the Spent Fuel Pool. There is however, a "SFP TROUBLE" lampbox 22M02 window F-6 annunciator in the control room. One of the inputs to this alarm is FC-LSHL-1420, the ICSA HI/LO level sensor. Since no fuel is located in the ICSA in modes 1-4, this EAL only applies in modes 5 or 6.Area radiation monitors RE-8055 and RE-8099 are located are located in the RCB 68' elevation on the bioshield wall close to the refueling cavity. Area radiation monitor RE-8090 is located in the Fuel Handling Building on 68' Elevation near the Spent Fuel Pool.Expected radiation levels for a loss of water level can range from a few mR/hr to thousands of R/hr.For a drop of water level of approximately 14' (from 66'-6" to 51 '-10") with approximately 13' of water over the top of any array, the dose rate would be expected not to exceed 2.5 mR/hr, above background. This assumes 42 hours of decay with a full core off-load (section 9 of STP UFSAR).For a significant drop of water level that would still cover the arrays, the radiation levels could range from several hundred R/hr to over a thousand R/hr on and around the 68' elevation deck (table C-5 NUREG CR/0649).

REFERENCES:

I. OPOP09-AN-22M2, Rev. 25, Annunciator Lampbox 22M02 Response Instructions F-5 and F-6 Window (level alarms)2. 0POP04-FC-0001, Rev. 29, Loss of Spent Fuel Pool Level or Cooling (level alarms)3. Technical Specification, amendment 104 (Unit 1) and 91 (Unit 2), Section 5.6.2 (Design water level)4. UFSAR, Rev. 16, Section 9.1.2.1 (Dose rates)5. UFSAR, Rev. 16, Section 9.1.2.2 (Normal water level)6. NUREG CR/0649 (Dose rates), reference only (not included in submittal)

7. Drawing 5R219F05028#1 Spent Fuel Pool Cooling and Cleanup System (level sensors)8. UFSAR, Rev. 15, table 12.3.4-1, Area Radiation Monitors 24 1 P a-, e RA1 ECL: ALERT Initiating Condition:

Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem THYROID CDE.Operating Mode Applicability: ALL Emergency Action Levels: (I or 2 or 3 or 4)Notes:* The Emergency Director should declare the ALERT promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.* The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification.assessments until the results from a dose assessment using actual meteorology are available. (1) Reading on ANY of the following radiation monitors greater than the values listed in Table RI column"ALERT" for 15 minutes or longer: Table RI: Effluent Monitors Release Point Monitor GE SAE ALERT UE Unit Vent RT-8010B 1.50 E+08 pCi/sec 1.50 E+07 pCi/sec 1.50 E+06 ptCi/sec 1.40 E+05 pCi/sec Main Steam RT-8046 thru 4.00 E+02 pCi/cm 3 4.00 E+01 gCi/cm 3 4.00 E+00 pLCilcrn 3 5.00 E-02 pCi/cm 3 Lines 8049 (2) Dose assessment using actual meteorology indicates doses greater than 10 mremn TEDE or 50 mrem THYROID CDE at or beyond the SITE BOUNDARY.(3) Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem THYROID CDE at or beyond the SITE BOUNDARY for one hour of exposure.(4) Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:* Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate THYROID CDE greater than 50 mrem for one hour of inhalation. 25 1P a e Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA PROTECTIVE ACTION GUIDES (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem THYROID CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and THYROID CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC RS1.RAI: EAL-1 Selection Basis The Unit Vent and Main Steam Line monitor readings were included in this EAL because they give instantaneous indications of a monitored gaseous release meeting the EAL threshold values of 10 mrem TEDE or 50 mrem CDE THYROID at the SITE BOUNDARY. Gaseous releases from the plant are monitored by the Unit Vent. The Main Steam Line monitor readings correspond to a concentration that would result in a release rate meeting the EAL threshold values if there were a release via the Power Operated Relief Valves (PORVs) or Safety Relief Valves.The Unit Vent and Main Steam Line release values are based on Calculation No. STPNOCOI3-CALC-002, Rev. 2. The adjusted values used in this EAL were conservatively truncated by less than 1% of the calculated values to ensure they are readily assessable. RAI: EAL-2, 3, 4 Selection Basis N/A

REFERENCES:

1. Calculation No: STPNOC013-CALC-002 Rev. 2., Radiological Release Thresholds for Emergency Action Levels 2. UFSAR, Rev. 14, Section 11.5.2.3.3 and 11.5.2.5.3 (monitor descriptions)

3. UFSAR, Rev. 14, 11.5.2.4.4 (liquid waste processing monitor)261 P a o e RA2 ECL: ALERT Initiating Condition:

Significant lowering of water level above, or damage to, irradiated fuel.Operating Mode Applicability: ALL Emergency Action Levels: (I or 2 or 3)(1) Uncovery of irradiated fuel in the REFUELING PATHWAY.(2)a. Damage to irradiated fuel resulting in a release of radioactivity friom the fuel as indicated by ANY of the following FHB radiation monitor readings:* FHB Exhaust, RT-8035 or RT-8036 greater than 1.00 E- I.tCi/cm 3 OR* ARM (68' FHB), RE-8090 greater than 1,500 mR/hr OR b. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of the following RCB radiation monitor readings (Mode 5 or 6 only).* ARMs (68' RCB), RE-8055 or RE-8099 greater than 850 mR/hr.NOTE EAL-3 is not applicable until the enhanced SFP level instrumentation is available for use.(3) Lowering of spent fuel pool level to 49'-10" or lower.Basis: This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fitel pool or Inside Containment Storage Area (UCSA).These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HUI.Escalation of the emergency would be based on either Recognition Category R or C ICs.271 P a ge EAL #1- This EAL escalates friom RU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in Uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations. While an area radiation monitor could detect a rise in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.EAL #2- This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel.Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident). EAL #3- Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventoly and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.Escalation of the EMERGENCY CLASSISFICATION LEVEL would be via ICs RSI or RS2.RA2: EAL-2 Selection Basis: The calculated airborne source term and radiation monitor responses for a fuel handling accident in the FHB is based on Calculation STPNOCOI3-CALC-005 Rev.2. The threshold value of 1500 mR/hr for area radiation monitor RE-8090 was truncated less than 4% from the calculated value to ensure the threshold was readily assessable. Threshold values for FHB Exhaust Monitors RT-8035 and RT-8036 were also included because they are accident monitors that are sensitive to noble gases which are expected to be present if irradiated fuel is damaged. The calculated monitor reading for RT-8035 and RT-8036 is 3.8 ýtCi/cm 3 and the high range of the monitors is 0.3 [tCi/cm 3.The threshold value of 0.1 pCi/cm 3 is approximately 6 orders of magnitude above background and indicative of damaged irradiated fuel. It was selected because it is readily assessable and within the calibrated range of the monitors.The calculated airborne source term and radiation monitor response for a fuel handling accident in the RCB is based on Calculation STPNOC013-CALC-005 Rev.2. The threshold value of 850 mR/hr for RE-8055 and RE-8099 was truncated less than 2% from the calculated value to ensure the threshold is readily assessable. RA2: EAL-3 Selection Basis: Spent Fuel Pool level of 49'- 10" (Level 2) is a site specific level based on the guidance provided in NEI 12-02, Revision 1, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licensees with Regard to Reliable Spent Fuel Pool Instrumentation", August 2012.In NRC Order EA-12-051 and NEI 12-02., Level 2 is defined as the "level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck ... " 28 1 P a g e The STP UFSAR identifies the top of the Spent Fuel Storage Racks at 39'- 10". The guidance in NEI 12-02 indicates that 10' of water above the top of the Spent Fuel Storage Racks provides substantial radiation shielding. Ten feet of water above the Spent Fuel Storage Racks is 49'- 10", the threshold value for this EAL.Reference 6 identifies the site specific levels of the proposed SFP level instrument and identifies the Level 2 criteria as 49'- 10".

REFERENCES:

1. Calculation No.: STPNOCOI3-CALC-005 Rev.2, Fuel Handling Accident Monitor Response for Emergency Action Levels.2. UFSAR, Rev. 16, Section 9.1.2.1 (SFP Rad levels)3. UFSAR, Rev. 16, Section 9.1.2.2 (SFP top of Racks)4. NRC Order EA-12-051 (SFP levels)5. NEI 12-02, Rev. 1 (SFP levels)6. South Texas Project (STP) Overall Integrated Plan for Implementation of Unit I & Unit 2 Spent Fuel Pool Level histrumentation to Meet NRC Order EA- 12-051, Rev. 0, NOC -AE- 13002959 29 1P a ge RA3 ECL: ALERT Initiating Condition:

Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.Operating Mode Applicability: ALL Emergency Action Levels: (1 or 2)Note: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. (1) Dose rate greater than 15 mR/hr in ANY of the following areas:* Control Room ARM (RE-8066)OR* Central Alarm Station (CAS) by radiation survey (2) An UNPLANNED event results in radiation levels that prohibit or impede access to ANY of the areas listed in Table H3/R2:__.___ ;TABLE H3/R2: Plant'Areas'l.Rquiring Access RCB RHR Heat Exchanger Rooms 0 0 MAB 51 ft Room 335 EAB Roof, MCC 1G8, 4.16KV Switchgear Rooms LU 0 0 "n EAB 4. 16KV Switchgear Rooms 30 1 P a g e Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the higher radiation levels and determine if another IC may be applicable. For EAL #2, an ALERT declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the higher radiation levels.Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).An emergency declaration is not warranted if any of the following conditions apply.* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode I when the radiation rise occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.* The higher radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).* The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

• The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via Recognition Category R, C or F ICs.RA3: EAL-1, EAL-2 Selection Basis: The NEI 99-01 value of 15 mR/hr is derived from the GDC 19 value of 5 rem in 30 days with adjustment for expected occupancy times. The rooms listed in EAL-1 require continuous occupancy to maintain normal plant operation, or to perform a normal cooldown or shutdown.The areas listed in EAL-2 apply to areas that contain equipment necessary for plant operations, cooldown, or shutdown.

Assuming all plant equipment is operating as designed, Normal operations and safe shutdown equipment operation is capable from the Main Control Room (MCR). The plant is able to transition into a hot shutdown from the MCR, therefore H3/R2 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary following entry into hot shutdown (establish Residual Heat Removal shutdown cooling, disable operation of charging and ECCS equipment, and limit dilution pathways) and subsequent entry into cold shutdown (disable operation of ECCS equipment). After achieving cold shutdown it is assumed that the plant will be maintained in a cold shutdown condition. 31 P a e

REFERENCES:

1. General Design Criteria 19 2. OPOP03-ZG-0008, Rev. 56, Power Operations

3. OPOP03-ZG-0006, Rev. 54, Plant Shutdown from 100% to Hot Standby 4. OPOP03-ZG-0007, Rev. 7 1, Plant Cooldown 32 1 P a ge RS1 ECL: SITE AREA EMERGENCY Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 100 rnrern TEDE or 500 mrem THYROID CDE.Operating Mode Applicability: ALL Emergency Action Levels: (I or 2 or 3)Notes: The Emergency Director should declare the SITE AREA EMERGENCY promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.* The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. (1) Reading on ANY of the following radiation monitors greater than the values listed in Table RI column"SAE" for 15 minutes or longer: Table RI: Effluent Monitors Release Point Monitor GE SAE ALERT UE Unit Vent RT-8010B 1.50 E+08 1.50 E+07 pCi/sec 1.50 E+06 1.40 E+05 ýtCi/sec Main Steam RT-8046 thru 4.00 E+02 pCi/cm 3 4.00 E+01 pCi/cm 3 4.00 E+00 ptCi/cm 3 5.00 E-02 pCi/cm 3 Lines 8049 (2) Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem THYROID CDE at or beyond the SITE BOUNDARY.(3) Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:* Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate THYROID CDE greater than 500 mrem for one hour of inhalation. 33' I P a ge Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA PROTECTIVE ACTION GUIDES (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that carmot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem THYROID CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and THYROID CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC RGI.RS1: EAL-1 Selection Basis: The Unit Vent and Main Steam Line monitor readings were included in this EAL because they give instantaneous indications of a monitored gaseous release meeting the EAL threshold values of 100 mrem TEDE or 500 mremn CDE THYROID at the SITE BOUNDARY. Gaseous releases from the plant are monitored by the Unit Vent. The Main Steam Line monitor readings correspond to a concentration that would result in a release rate meeting the EAL threshold values if there were a release via the Power Operated Relief Valves (PORVs) or Safety Relief Valves.The Unit Vent and Main Steam Line release values are based on Calculation No. STPNOC013-CALC-002 Rev.2. The adjusted values used in this EAL were conservatively truncated by less than 1% of the calculated values to ensure they are readily assessable. RS1: EAL-2. EAL-3 Selection Basis: N/A

REFERENCES:

1. Calculation No: STPNOCO I 3-CALC-002 Rev.2, Radiological Release Thresholds for Emergency Action Levels 2. UFSAR Section, Rev. 14, Section 11.5.2.3.3 and 11.5.2.5.3 (monitor descriptions) 34 1 P a g e RS2 ECL: SITE AREA EMERGENCY Initiating Condition:

Spent fuel pool level at 40'-4" or lower.Operating Mode Applicability: ALL Emergency Action Level: NOTE EAL-I is not applicable until the enhanced SFP level instrumentation is available for use.(1) Lowering of spent fuel pool level to 40'-4" or lower.Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a SITE AREA EMERGENCY declaration. It is recognized that this IC would likely not be met until well after another SITE AREA EMERGENCY IC was met; however, it is included to provide classification diversity. Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC RGI or RG2.RS2: EAL-1 Selection Basis: Spent Fuel Pool level of 40'- 4" (Level 3) is a site specific level based on the guidance provided in NEI 12-02, Revision 1, Industry Guidance for Compliance with NRC Order EA-1 2-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", August 2012.In NRC Order EA-1 2-051 and NEI 12-02, Level 3 is defined as "level where fuel remains covered and actions to implement make-up water addition should no longer be deferred. " The STP UFSAR identifies the top of the Spent Fuel Storage Racks at 39'- 10".Reference 4 identifies the site specific levels for the proposed SFP level instrumentation and identifies the Level 3 criteria as 40'- 4".

REFERENCES:

1. UFSAR, Rev. 16, Section 9.1.2.2 (SFP top of Racks)2. NRC Order EA- 12-051 (SFP Levels)3. NEl 12-02, Revision 1, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", August 2012 1. South Texas Project (STP) Overall Integrated Plan for Implementation of Unit I & Unit 2 Spent Fuel Pool Level Instrumentation to Meet NRC Order EA-12-051, Rev. 0, NOC -AE-13002959 35 I P a ,, e RG1 ECL: GENERAL EMERGENCY Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrern THYROID CDE.Operating Mode Applicability: ALL Emergency Action Levels: (1 or 2 or 3)Notes:* The Emergency Director should declare the GENERAL EMERGENCY promptly upon determining that the applicable time has been exceeded, or will likely be exceeded." If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.* The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. (1) Reading on ANY of the following radiation monitors greater than the values listed in Table RI column"GE" for 15 minutes or longer: Table RI: Effluent Monitors Release Point Monitor GE SAE ALERT LIE Unit Vent RT-8010B 1.50 E+08 iCi/sec 1.50 E+07 piCi/sec 1.50 E+06 LCi/sec 1.40 E+05 ICi/sec Main Steam RT-8046 thni 4.00 E+02 iLCi/cm 3 4.00 E+01 piCi/cm 3 4.00 E+00 lpCi/cm 3 5.00 E-02 pCi/cm 3 Lines 8049 (2) Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem THYROID CDE at or beyond the SITE BOUNDARY.(3) Field survey results indicate EITHER of the following at or the SITE BOUNDARY:* Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.OR* Analyses of field survey samples indicate THYROID CDE greater than 5,000 mrem for one hour of inhalation. 36 1P ag-e Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA PROTECTIVE ACTION GUIDES (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem THYROID CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and THYROID CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.RG1: EAL-1 Selection Basis: The Unit Vent and Main Steam Line monitor readings were included in this EAL because they give instantaneous indications of a monitored gaseous release meeting the EAL threshold values of 1000 mrem TEDE or 5000 mrem CDE THYROID at the SITE BOUNDARY. Gaseous releases from the plant are monitored by the Unit Vent. The Main Steam Line monitor readings correspond to a concentration that would result in a release rate meeting the EAL threshold values if the release was via the Power Operated Relief Valves (PORVs) or Safety Relief Valves.The Unit Vent and Main Steam Line release values are based on Calculation No. STPNOCOI3-CALC-002 Rev.2. The adjusted values used in this EAL were conservatively truncated by less than 1% of the calculated values to ensure they are readily assessable. RGI: EAL-2. EAL-3 Selection Basis: N/A

REFERENCES:

1. Calculation No: STPNOCO13-CALC-002 Rev.2, Radiological Release Thresholds for Emergency Action Levels, 2. STP UFSAR, Rev. 14, Section 11.5.2.3.3 and 11.5.2.5.3 (monitor descriptions) 371 P a ge RG2 ECL: GENERAL EMERGENCY Initiating Condition:

Spent fuel pool level cannot be restored to at least 40'-4" for 60 minutes or longer.Operating Mode Applicability: ALL Emergency Action Level: Note: The Emergency Director should declare the GENERAL EMERGENCY promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.NOTE EAL-1 is not applicable until the enhanced SFP level instrumentation is available for use.(1) Spent fuel pool level cannot be restored to at least40'-4" for 60 minutes or longer.Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment. It is recognized that this IC would likely not be met until well after another GENERAL EMERGENCY IC was met; however, it is included to provide classification diversity. RG2: EAL-1 Selection Basis: The Spent Fuel Pool level of 40'- 4" (Level 3) is a site specific level based on the guidance provided in NEI 12-02, Revision 1, Industry Guidance for Compliance with NRC Order EA-12-05 1, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", August 2012.In NRC Order EA-l 2-051 and NEI 12-02, Level 3 is defined as "level where fuel remains covered and actions to implement make-up water addition should no longer be deferred. " The STP UFSAR identifies the top of the Spent Fuel Pool Racks at 39'- 10".Reference 4 identifies the site specific levels of the proposed level instrumentation and identifies the Level 3 criteria as 40'- 4".381 P ag e

REFERENCES:

1. UFSAR, Rev. 16, Section 9.1.2.2 (SFP top of Racks)2. NRC Order EA-12-051 (SFP Levels)3. NEI 12-02, Rev. 1, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", August 2012 4. South Texas Project (STP) Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Level Instrumentation to Meet NRC Order EA-12-051, Rev. 0, NOC -AE-13002959 3911-1ag e 7 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS Table C-i: Recognition Category "C" Initiating Condition Matrix UNUSUAL EVENT CU1 UNPLANNED loss of RCS inventory for 15 minutes or longer.Op. Modes: 5,6 CU2 Loss of ALL but one AC power source to emergency buses for 15 minutes or longer.Op. Modes: 5,6 Defueled CU3 UNPLANNED rise in RCS temperature.

Op. AModes: 5,6 CU4 Loss of Vital DC power for 15 minutes or longer.Op. Modes: 5,6 ALERT SITE AREA EMERGENCY GENERAL EMERGENCY CA1 Loss of RCS inventory. Op. Modes. 5,6 CS1 Loss of RCS inventory affecting core decay heat removal capability. Op. Modes: 5,6 CG1 Loss of RCS inventory affecting fuel clad integrity with containment challenged. Op. Modes: 5,6 CA2 Loss of ALL offsite and ALL onsite AC power to emergency buses for 15 minutes or longer.Op. Modes: 5,6, Defueled CA3 Inability to maintain the plant in cold shutdown.Op. Modes: 5,6 CU5 Loss of ALL onsite or offsite communications capabilities. Op. Modes: 5,6, Defuieled CA6 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.Op. Modes: 5,6 40 Pa e cul ECL: UNUSUAL EVENT Initiating Condition: UNPLANNED loss of RCS inventory for 15 minutes or longer.Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) UNPLANNED loss of reactor coolant results in RCS level below the procedurally required limit for 15 minutes or longer.(2) a. RCS level cannot be monitored. AND b. UNPLANNED rise in ANY of the following sump or tank levels in Table C2: Table C2: RCS Leakage* Containment Normal Sump* Pressurizer Relief Tank (PRT)* Reactor Coolant Drain Tank (RCDT)* MAB Sumps I thru 4* Containment Penetration Area Sump* SIS/CSS Pump Compartment Sump Basis: This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RCS level concurrent with indications of coolant leakage.Either of these conditions is considered to be a potential degradation of the level of safety of the plant.Refueling evolutions that lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an UNUSUAL EVENT due to the reduced water inventory that is available to keep the core covered.EAL #1- recognizes that the minimum required RCS level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is specified in the applicable STP operating procedure. The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.411 Pa- e EAL #2- addresses a condition where all means to determine RCS level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels.Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS Continued loss of RCS inventory may result in escalation to the ALERT EMERGENCY CLASSIFICATION LEVEL via either IC CA 1 or CA3.CUI -EAL-1 Selection Basis: RCS inventory is maintained above the reactor vessel flange (39'-3") during refueling outages per OPOP03-ZG-0007, Plant Cooldown. RCS level may be lowered below the vessel flange for specific purposes (e.g., head removal, mid-loop operations) as described in OPOP03-ZG-0009, Mid-Loop Operation. The 15 minute time frame allows for prompt operator actions to restore RCS level in the event of an UNPLANNED lowering of RCS level below the prescribed operating limit.CU1 -EAL-2 Selection Basis: This EAL includes two conditions. The first condition is the inability to monitor RCS level and the second condition provides secondary indications that inventory loss may be occurring. The secondary indicators of inventory loss include a list of tanks/sumps found in OPOP04-RC-0003, Excessive RCS Leakage. Since other system leaks could rise levels in various tanks and sumps, the list has been limited to the tanks and sumps that would have the highest probability of indicating RCS leakage inside the Reactor Containment Building.Although procedure OPOP04-RC-0003 is designated for use in modes 1-4, its logic is applicable to this EAL.

REFERENCES:

1. OPOP04-RC-0003, Rev. 18, Excessive RCS Leakage 2. OPOP03-ZG-0007, Rev. 71, Plant Cooldown 3. OPOP03-ZG-0009, Rev. 59, Mid-Loop Operation 42 1 Page CU2 ECL: UNUSUAL EVENT Initiating Condition:

Loss of ALL but one AC power source to emergency buses for 15 minutes or longer.Operating Mode Applicability: 5, 6, Defueled Emergency Action Level: Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. AC power capability to ALL three 4160V AC ESF Buses is reduced to a single power source for 15 minutes or longer.AND b. ANY additional single power source failure will result in loss of ALL AC power to SAFETY SYSTEMS.Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as all ALERT because of the additional time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Examples of this condition are presented below.* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

• A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from the unit main generator.
• A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from an onsite or offsite power source.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.Tile subsequent loss of the remaining single power source would escalate the event to an ALERT in accordance with IC CA2.431 P a oe CU2: EAL-1 Selection Criteria: The condition indicated by this EAL is the degradation of the offsite and onsite power systems such that any additional single failure would results in a loss of all AC power. This condition is an UNUSUAL EVENT during modes 5, 6 and Deftieled because of the additional time available to restore power due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. In modes 1-4, this condition is an ALERT as described in SAL.

REFERENCES:

1. OPOP04-AE-000 1 Rev. 44. First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus 2. OPOP04-AE-0004, Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus 3. OPSP03-EA-0002, Rev. 32, ESF Power Availability

4. Drawing OOOOOEOAAAA, Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. 1 & 2 44 1 P a g e CU3 ECL: UNUSUAL EVENT Initiating Condition:

UNPLANNED rise in RCS temperature. Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) UNPLANNED rise in RCS temperature to greater than 200 'F (Tavg).(2) Loss of ALL RCS temperature and RCS level indication for 15 minutes or longer.Basis: This IC addresses an UNPLANNED rise in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA3.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. EAL #1- involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange.Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid rise in reactor coolant temperature depending on the time after shutdown.EAL #2- reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation to ALERT would be via IC CAl based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.451 P a ge CU3: EAL-I Selection Basis: An UNPLANNED temperature rise above 200 'F would result in an UNPLANNED mode change due to the inability to control RCS temperature. Mode 4 (Hot Shutdown) would be entered when Tavg exceeds 200 'F (Reference 1).CU3: EAL-2 Selection Basis: N/A

REFERENCES:

1. Technical Specifications Table 1.2 (Mode, Temperature, Power, ketf Table)46 1 P a e CU4 ECL: UNUSUAL EVENT Initiating Condition:

Loss of Vital DC power for 15 minutes or longer.Operating Mode Applicability: 5, 6 Emergency Action Level: Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Indicated voltage is less than 105.5 VDC on required Vital DC buses for 15 minutes or longer.Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions extend the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.As used in this EAL, "required" means the Vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A and C are out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an UNUSUAL EVENT. A loss of Vital DC power to Train A and/or C would not warrant an emergency classification. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Depending upon the event, escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC CAI or CA3, or an IC in Recognition Category R.CU4 -EAL-1 Selection Basis: The minimum voltage for Class 1E 125 VDC battery buses was determined in calculation 13-DJ-006, Rev. 3 to be 105.5 volts. At 105.5 volts or less, OPOP05-EO-EC0O, Loss of All AC Power, directs the operators to open the battery output breakers.

REFERENCES:

1. Calculation 13-DJ-006, Rev. 0, 125 VDC Battery Four Hour Coping Analysis 2. OPOP05-EO-ECOO, Rev. 23, Loss of All AC Power 47 1 P a g e CU5 ECL: UNUSUAL EVENT Initiating Condition:

Loss of ALL onsite or offsite communications capabilities. Operating Mode Applicability: 5, 6, Deflieled Emergency Action Levels: (1 or 2 or 3)(1) Loss of ALL of the following Onsite communication methods in Table C4.(2) Loss of ALL of the following Offsite Response Organization (ORO) communication methods in Table C4.(3) Loss of ALL of the following NRC communication methods in Table C4.Table C4: Communications Methods EAL-1 EAL-2 EAL-3 ONSITE ORO NRC Plant PA system X Plant Radios X Plant telephone system X X X Satellite phones X X Direct line from Control Rooms to Bay City X X Microwave Lines to Houston X X Security radio to Matagorda County X Dedicated Ring-down lines X ENS line X Basis: This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).EAL #1-addresses a total loss of the communications methods used in support of routine plant operations. EAL #2-addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are Matagorda County Sheriff's Office. and Texas Department of Public Safety Disaster District in Pierce.481 P a g e EAL #3-addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. CU5: EAL-1. EAL-2. and EAL-3 Selection Basis: Lines not included for offsite comLmunications to ORO and NRC included links that would need relaying of information. Links were obtained from procedures OPGP05-ZV-001 L, Emergency Communications.

REFERENCES:

1. OPGP05-ZV-001 1, Rev. 8, Emergency Communications 491 P a g e CA1 ECL: ALERT Initiating Condition:

Loss of RCS inventory. Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the ALERT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Loss of RCS inventory as indicated by level less than 32 ft. 9 inch (+ 6 inches above hot leg centerline). (2) a. RCS level cannot be monitored for 15 minutes or longer AND b. UNPLANNED rise in ANY of the following sump or tank levels in Table C2 due to a loss of reactor vessel/RCS inventory. Table C2: RCS Leakage* Containment Normal Sump* Pressurizer Relief Tank (PRT)* Reactor Coolant Drain Tank (RCDT)* MAB Sumps I thru 4* Containment Penetration Area Sump* SIS/CSS Pump Compartment Sump Basis: This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.EAL #1- A lowering of water level below elevation 32'- 9" indicates that operator actions have not been successful in restoring and maintaining reactor vessel/ water level. The heat-up rate of the coolant will rise as the available water inventory is reduced. A continuing reduction in water level will lead to core uncovery. Although related, EAL #1 is concerned with the loss of RCS inventory and.not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Heat Removal suction point). Arise in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.EAL #2- The inability to monitor reactor vessel/RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to 50 P a e ensure they are indicative of leakage from the reactor vessel/RCS. The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CSI If the reactor vessel/RCS inventory level continues to lower, then escalation to SITE AREA EMERGENCY would be via IC CS 1.CAl: EAL-1 Selection Basis: The minimum RCS level at which an RHR pump can be started per OPOP02-RH-000 I is 32 feet 9 inches (+ 6 inches above hot leg centerline). If RCS inventory is reduced below this level, normal decay heat removal systems may not be available for core cooling. This threshold is not applicable to reduced inventory vacuum fill since this is a controlled evolution and not indicative of an RCS loss.CAI: EAL-2 Selection Basis: The tanks/sumps selected for this EAL were obtained from OPOP04-RC-0003, Excessive RCS Leakage. Since other system leaks could raise levels in various tanks and sumps, the list was limited to the tanks and sumps that would have the highest probability of indicating RCS leakage inside the Reactor Containment Building.Although procedure OPOP04-RC-0003 is designated for use in modes 1-4, its logic is applicable to this EAL.

REFERENCES:

I. OPOP04-RC-0003, Rev. 18, Excessive RCS Leakage 2. OPOP02-RH-00011, Rev. 63, Residual Heat Removal System Operation 51 1Page CA2 ECL: ALERT Initiating Condition: Loss of ALL offsite and ALL onsite AC power to emergency buses for 15 minutes or longer.Operating Mode Applicability: 5, 6, Defuieled Emergency Action Level: Note: The Emergency Director should declare the ALERT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Loss of ALL offsite AND ALL onsite AC Power to ALL three 4160V AC ESF Busses for 15 minutes or longer.Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a SITE AREA EMERGENCY because of the additional time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC CSI or RS1.CA2 -EAL-1 Selection Basis: N/A

REFERENCES:

1. OPOP04-AE-0001, Rev. 44, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus 2. OPOP04-AE-0004, Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus 3. OPSP03-EA-0002, Rev. 32, ESF Power Availability

4. Drawing OOOOOEOAAAA, Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. I & 2 52 111 a ge CA3 ECL: ALERT Initiating Condition:

Inability to maintain the plant in cold shutdown.Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the ALERT promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.(1) UNPLANNED rise in RCS temperature to greater than 200 'F (Tavg) for greater than the duration specified in Table C3.Table C3: RCS Heat-up Duration Thresholds RCS Status Containment Closure Status Heat-up Duration Intact (but not at reduced inventory) Not applicable 60 minutes*Not intact (or at reduced inventory) Established 20 minutes*Not Established 0 minutes* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable. (2) UNPLANNED RCS pressure rise greater than 10 psig. (This EAL does not apply during water-solid plant conditions.) Basis: This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.A momentary UNPLANNED excursion above the Teclmical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. EAL #1-The RCS Heat-up Duration Thresholds table addresses an rise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation). The 20-minute criterion was included to allow time for operator action to address the temperature rise.The RCS Heat-up Duration Thresholds table also addresses an rise in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature rise without a substantial degradation in plant safety.531 P a ge Finally, in the case where there is a rise in RCS temperature, the RCS is not intact or is at reduced inventory and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.EAL #2- provides a pressure-based indication of RCS heat-up.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC CS1 or RSI.CA3 -EAL-1 Selection Basis: Table C3 was adopted from NEI 99-01, Rev. 6. This EAL addresses the concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal. A number of phenomena such as pressurization, vortexing, steam generator U-tube draining, RCS level differences when operating at a mid-loop condition, decay heat removal system design, and level instrumentation problems can lead to conditions where decay heat removal is lost and core uncover can occur. NRC analyses show that there are sequences that can cause core uncovery in 15 to 20 minutes, and severe core damage within an hour after decay heat removal is lost. The allowed time frames are consistent with the guidance provided by Generic Letter 88-17 and believed to be conservative given that a low pressure containment barrier to fission product release is established. CA3 -EAL-2 Selection Basis: An UNPLANNED RCS pressure rise greater than 10 psig provides a pressure-based indication of RCS heat-up.The pressure change, per NEI 99-01 Rev. 6, is the lowest change in pressure that can be accurately determined using installed instrumentation, but not less than 10 psig.

REFERENCES:

I. Technical Specifications Table 1.2 (Mode, Temperature, Power., keff Table)5411Page CA6 ECL: ALERT Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.Operating Mode Applicability: 5, 6 Emergency Action Level: (1) a. The occurrence of ANY of the following hazardous events in Table C5: Table C5: Hazardous Events" Seismic event (earthquake)

• Internal or external flooding event" High winds or tornado strike" FIRE" EXPLOSION* Predicted or actual breach of Main Cooling Reservoir retaining dike along the North Wall* Other events with similar hazard characteristics as determined by the Shift Manager AND b. EITHER of the following:

1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.OR 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.Basis: This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.EAL#1.b.1-addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.

The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.EAL#I.b.2 addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will 55 1r a -e make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC CS1 or RSI.CA6: EAL-1 Selection Basis: The listed hazards are taken directly from NEI 99-0 1, Rev. 6. The only additional hazard was the inclusion of the Main Cooling Reservoir since it is a credible hazard and analyzed in the STPEGS UFSAR (reference 2).

REFERENCES:

I. STPEGS UFSAR, Rev. 13, Section 3.4.1, Flood Protection 56 P1 a g e CS1 ECL: SITE AREA EMERGENCY Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability. Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2 or 3)Note: The Emergency Director should declare the SITE AREA EMERGENCY promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.(1) a. CONTAINMENT CLOSURE not established. AND b. RCS level less than 33% of plenum.(2) a. CONTAINMENT CLOSURE established. AND b. RCS level less than 0% of plenum (3) a. RCS level cannot be monitored for 30 minutes or longer.AND b. Core uncovery is indicated by ANY of the following:

• Reactor Containment Building, 68'-0" Area Radiation Monitors RE-8055 or RE-8099 reading greater than 9,000 mR/hr.OR* Erratic source range monitor indication.

OR" UNPLANNED rise in ANY of the following sump or tank levels in Table C2 of sufficient magnitude to indicate core uncovery.Table C2: RCS Leakage* Containment Normal Sump* Pressurizer Relief Tank (PRT)* Reactor Coolant Drain Tank (RCDT)* MAB Sumps 1 thru 4* Containment Penetration Area Sump* SIS/CSS Pump Compartment Sump 57 1 P a g e Basis: This IC addresses a significant and prolonged loss of reactor vessel/RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a SITE AREA EMERGENCY declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in RCS level. If RCS level cannot be restored, fuel damage is probable.Outage/shutdown contingency plans provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS levels of EALs L.b and 2.b reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the envirormnent. In EAL 3.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurTed (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issutes; NUREG-1 449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Mlanagement. Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC CG I orRG 1.CS1: EAL-1 Selection Basis: Per NEI 99-01 Rev. 6, the RCS level indication should be six inches (6") below the bottom inside diameter of the RCS loop penetration at the reactor vessel. Six inches (6") below the bottom inside diameter of the RCS hot leg nozzle (elevation 31 '-0.5") is elevation 30'-6.5" per 0POP03-ZG-0009, Mid-Loop Operation, Addendum 1, RCS/RHR Simplified Elevation Diagram. The nearest RVWL Monitoring System thermocouples are located 6 inches above (Sensor 6) and 4.9 inches below (Sensor7) the prescribed elevation of 30'-6.5". When water level is at the desired elevation of 30'-6.5", Sensor 6 will be dry and Sensor 7 will be wet. This condition corresponds to a reading of 33% of plenum per 0POP02- 11-0002, RVWL Monitoring System, Addendum 1, RVWL Sensor Elevations. 581 P ag e CS1: EAL-2 Selection Basis: Per NEI 99-01 Rev. 6, the RCS level indication should be approximately the top of active fuel (TAF). The RCS level which corresponds to the top of the active fuel is 28'-2" (0POP03-ZG-0009, Mid-Loop Operation, Addendum 1, RCS/RHR Simplified Elevation Diagram). The nearest Reactor Vessel Water Level Monitoring System thennocouple to TAF is Sensor 8 at elevation 29'-2.7". Use of RVWL to approximate TAF; with the inherent gap of 12 inches between indicated level and actual level, is acceptable for the purposes of signaling that the threat to the public is reduced when CONTAINMENT CLOSURE is established. CS1: EAL-3 Selection Basis: As RCS level drops the dose rates above the core will rise. Area Radiation Monitors RE-8055 and RE-8099 are located on the 68'-0" elevation of the reactor containment building. Their locations are identified on drawing 9C129A81105. Their range (0.1 mR/hr to 10,000 mR/hr) is identified in Table 12.3.4-1 of Section 12 of the UFSAR. A rising trend on these monitors can be an indication that core uncovery is occurring. Additionally, erratic source range monitor indications, or large level rises in the tanks listed can give further indication of core uncovery.The threshold value for radiation monitors RE-8055 and RE-8099 was based on Calculation STPNOC013-CALC-006 Rev.2. The calculated monitor response is 22.4 R/hr when RCS level is at the top of the active fuel and 6 R/hr at one foot above the top of active fuel. The high range of these monitors is 10 R/hr. The value of 9,000 mR/hr was selected to ensure that the threshold is readily assessable and within the calibrated range of the monitor. The threshold value of 9,000 mnR/hr corresponds to approximately 8 inches above the top of the active fuel with the reactor head on; which provides an additional indication that RCS levels are near the point of fuel uncovery. These monitor readings in conjunction with the other threshold values allow for an accurate assessment of the EAL.Core uncovery can be determined by the secondary indications listed in this EAL. The secondary indicators of inventory loss include a list of tanks/sumps found in OPOP04-RC-0003, Excessive RCS Leakage. Since other system leaks could raise levels in various tanks and sumps, the list has been limited to the tanks and sumps that would have the highest probability of indicating RCS leakage inside the Reactor Containment.

REFERENCES:

I. Calculation No: STPNOCOI3-CALC-006 Rev.2, Dose Rate Evaluation of Reactor Vessel Water Levels during Refueling for EAL Thresholds

2. OPOP03-ZG-0009, Rev. 59, Mid-Loop Operation, Addendum 1, RCS/RHIR Simplified Elevation Diagram 3. USFAR, Rev. 15, Chapter 12, Table 12.3.4-1 4. OPOP02-11-0002, Rev. 15, RVWL Monitoring System 5. OPOP04-RC-0003, Rev 18, Excessive RCS Leakage 6. Drawing 9C129A81105, Re. 3, Radiation Zones, Reactor Containment Building, Plan at E. 68' -0" 59 1 Page CG1 ECL: GENERAL EMERGENCY Initiating Condition:

Loss of RCS inventory affecting fuiel clad integrity with containment challenged. Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the GENERAL EMERGENCY promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.(1) a. RCS level less than 0% of plenum for 30 minutes or longer.AND b. ANY indication from the Table C1.(2) a. RCS level cannot be monitored for 30 minutes or longer.AND b. Core uncovery is indicated by ANY of the following: " Reactor Containment Building, 68'-0" Area Radiation Monitors RE-8055 or RE-8099 reading greater than 9,000 mR/hr.OR" Erratic source range monitor indication OR* UNPLANNED rise in ANY of the following sump or tank levels in Table C2 of sufficient magnitude to indicate core uncovery AND c. ANY indication from Table C1 Table Cl: Containment Challenge" CONTAINMENT CLOSURE not established

• • >4% hydrogen exists inside containment
• UNPLANNED rise in containment pressure* IF CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, THEN declaration of a General Emergency is not required.60 1 P a g e Table C2: RCS Leakage* Containment Normal SulMp* Pressurizer Relief Tank (PRT)" Reactor Coolant Drain Tank (RCDT)* MAB Sumps I thru 4* Containment Penetration Area Sump" SIS/CSS Pumnp Compartment Sump Basis: This IC addresses the inability to restore and maintain RCS level above the top of active fuel with containment challenged.

This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in RCS level. If RCS level cannot be restored, fuel damage is probable.With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a GENERAL EMERGENCY is not required.The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to Support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use indications in Table Cl to assess whether or not containment is challenged. In EAL 2.b, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS 61 P age These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG- 1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industrjy Actions to Assess Shutdown Management. CG1: EAL-1 Selection Basis: Per NEI 99-01 Rev. 6, the RCS level indication should be approximately the top of active fuel (TAF). The RCS level which corresponds to the top of the active fuel is 28'-2" (OPOP03-ZG-0009, Mid-Loop Operation, Addendum 1, RCS/RHR Simplified Elevation Diagram). The nearest Reactor Vessel Water Level Monitoring System thermocouple to TAF is Sensor 8 at elevation 29'-2.7". Use of RVWL to approximate TAF; with the inherent gap of 12 inches between indicated level and actual level, is acceptable for the purposes of maintaining the escalation logic for the loss of RCS level condition. CGI: EAL-2 Selection Basis: The secondary indicators of inventory loss include a list of tanks/sumps found in 0POP04-RC-0003, Excessive RCS Leakage. Since other system leaks could rise levels in various tanks and sumps, the list has been limited to the tanks and sumps that would have the highest probability of indicating RCS leakage inside the Reactor Containment Building.As RCS level drops the dose rates above the core will rise. Area Radiation Monitors RE-8055 and RE-8099 are located on the 68'-0" elevation of the reactor containment building. Their locations are identified on drawing 9C129A81105. Their range (0.1 mR/hr to 10,000 mR/hr) is identified in Table 12.3.4-1 of Section 12 of the UFSAR. Rises on these monitors can be can be an indication that core uncover is occurring. Additionally, erratic source range monitor indications, or large level rises in the tanks listed can give further indication of core uncovery.The threshold value for radiation monitors RE-8055 and RE-8099 was based on Calculation STPNOC013-CALC-006 Rev. 2. The calculated monitor response is 22.4 R/hr when RCS level is at the top of the active fuel and 6 R/hr at one foot above the top of active fuel. The high range of these monitors is 10 R/hr. The value of 9,000 mR/hr was selected for this threshold to ensure the threshold is readily assessable and within the calibrated range of the monitor. The threshold value of 9,000 lnR/hr with the reactor head on corresponds to approximately 8 inches above the top of the active fuel which provides an additional indication that RCS levels are near the point of fuel uncovery. These monitor readings in conjunction with the other threshold values allow for an accurate assessment of the EAL.

REFERENCES:

I. Calculation No. STPNOC013-CALC-006 Rev.2, Dose Rate Evaluation of Reactor Vessel Water Levels during Refueling for EAL Thresholds

2. OPOP03-ZG-0009, Rev. 59, Mid-Loop Operations

3. Drawing 9C129A81105, Rev. 3, Radiation Zones, Reactor Containment Building Plan at El. 68'-0" 4. USFAR, Rev. 15, Chapter 12, Table 12.3.4-1, Area Radiation Monitors 5. OPOP05-EO-E010, Rev. 21, Loss of Reactor or Secondary Coolant 6. OPOP04-RC-0003, Rev. 18, Excessive RCS Leakage 62 1 P a e 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)ICS/EALS Table E-1: Recognition Category "E" Initiating Condition Matrix UNUSUAL EVENT E-HU1 Damage to a loaded cask CONFINEMENT BOUNDARY.Op. Modes: ALL 631 P a ge E-HU1 ECL: UNUSUAL EVENT Initiating Condition:

Damage to a loaded cask CONFINEMENT BOUNDARY Operating Mode Applicability: ALL Emergency Action Level: (1) Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than: : a. 60 mrem/hr (gamma + neutron) on the top surface of the spent fuel cask OR b. 600 mreni/hr (gammna + neutron) onl the side surface of the spent fuel cask OR b. 7000 mrem/hr (gamma + neutron) on the side surface of the transfer cask.Basis: This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.The existence of "damage" is determined by radiological survey. The values for this EAL are 2 times the Technical Specification allowable radiation levels. The technical specification multiple of "2 times", which is also used in Recognition Category R IC RUI. is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.Security-related events for ISFSIs are covered under ICs HUI and HAI.E-HU1 -EAL-1 Selection Basis: NEI 99-01 Rev.6 states that the dose rate limits are 2 times the Cask Technical Specification Limits. Section 5.3.2 of the "Certificate of Compliance No. 1032, Appendix A, Technical Specifications For The HI-STORM FW MPC Storage System", states: 5.3.4 Notwithstandizg the limits established in Section 5.3.3, the measured close rates on a loaded OVERPACK or TRANSFER CASK shall not exceed the Jbllowing values: a. 30 mrem/hr (gamma + neutron) on the top qf the OVERPACIC 641 P a e

b. 300 mrem/hr (gamma + neutron) on the side of the OVERPACK.excluding inlet and outlet ducts c. 3500 mrem/hr (gamma + neutron) on the side of the TRANSFER CASK

REFERENCES:

1. Certificate of Compliance no. 1032, Appendix A, Technical Specifications For The HI-STORM FW MPC Storage System, Section 5.3, Radiation Protection Program.10 CFR 72.104, Criteria For Radioactive Materials In Effluents And Direct Radiation From An ISFSI or MRS 65 1 P a g e 9 FISSION PRODUCT BARRIER ICS/EALS Table 9-F-I: Recognition Category "F" Initiating Condition Matrix ALERT FA1 ANY Loss or ANY Potential Loss of either the Fuel Clad or RCS barrier.tOp. Modes: 1,2,3,4 SITE AREA EMERGENCY FS1 Loss or Potential Loss of ANY two barriers.Op. Modes: 1,2,3,4 GENERAL EMERGENCY FG1 Loss of ANY two barriers and Loss or Potential Loss of the third barrier.Op. Modes: 1,2,3,4 661 P ag e Table 9-F-2: EAL Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FA1 ALERT FS1 SITE AREA EMERGENCY FG1 GENERAL EMERGENCY ANY Loss or ANY Potential Loss of either the Fuel Loss or Potential Loss of ANY two barriers.

Loss of ANY two barriers and Loss or Potential Clad or RCS barrier. Loss of the third barrier.Fuel Clad Barrier RCS Barrier Containment Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 1. RCS or SG Tube Leakage 1. RCS or SG Tube Leakage 1. RCS or SG Tube Leakage Not Applicable A. Core Cooling -Orange A. An automatic or A. Operation of a standby A. A leaking or Not Applicable entry conditions met manual ECCS (SI) charging pump is RUPTURED SG actuation is required required by EITHER is FAULTED by EITHER of the of the following: outside of following: containment.

1. UNISOLABLE

1. UNISOLABLE RCS leakage RCS leakage OR OR 2. SG tube leakage.2. SG tube RUPTURE.B. Integrity

-Red entry conditions met 67 1 P a g e Fuel Clad Barrier RCS Barrier Containment Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 2. Inadequate Heat Removal 2. Inadequate Heat Removal 2. Inadequate Heat Removal A. Core Cooling -Red A. Core Cooling -Not Applicable A. Heat Sink -Red Not Applicable A. Core Cooling- Red entry conditions met Orange entry entry conditions entry conditions met conditions met met. for 15 minutes or OR longer.B. Heat Sink- Red entry conditions met 3. RCS Activity / Containment Radiation

3. RCS Activity / Containment Radiation

3. RCS Activity / Containment Radiation Al. RCB Rad Monitor Not Applicable A. Not Applicable Not Applicable Not Applicable Al. RCB Rad Monitor RT-8050 or RT- RT-8050 or RT-8051 greater than 8051 greater than 40 R/hr 380 R/hr OR OR 2. HATCH 2. HATCH MONITOR MONITOR greater than 90 greater than 840 mR/hr mR/hr OR B. Sample analysis indicates that reactor coolant activity is greater than 300 jiCi/gm dose equivalent 1- 13 1.68 1 P a g e Fuel Clad Barrier RCS Barrier Containment Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 4. Containment Integrity or Bypass 4. Containment Integrity or Bypass 4. Containment Integrity or Bypass Not Applicable Not Applicable Not Applicable Not Applicable A. Containment A. Containment

-Red isolation is required entry conditions met AND EITHER of OR the following: B. Explosive mixture 1. Containment exists inside integrity has containment been lost based (H 2 > 4%)on Emergency OR Director CI. Containment judgment. pressure greater OR than 9.5 psig.AND 2. UNISOLABLE pathway from 2. Less than one full the containment train of to the Containment Spray environment is operating per exists. design for 15 minutes or longer.OR B. Indications of RCS leakage outside of containment. 69 I P a g e Fuel Clad Barrier RCS Barrier Containment Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 5. Other Indications

5. Other Indications

5. Other Indications A. N/A A. N/A A. N/A A. N/A A. N/A A. N/A 6. Emergency Director Judgment 6. Emergency Director Judgment 6. Emergency Director Judgment A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the Emergency Director Emergency Director Emergency Director Emergency Director Emergency Director Emergency Director that indicates Loss that indicates that indicates Loss that indicates that indicates Loss that indicates of the Fuel Clad Potential Loss of the of the RCS Barrier. Potential Loss of the of the Containment Potential Loss of the Barrier. Fuel Clad Barrier. RCS Barrier. Barrier. Containment Barrier.70 1 P a g e Basis Information For EAL Fission Product Barrier Table 9-F-2 STP is part of the Westinghouse Owners Group (WOG) and has adopted the WOG Emergency Response Guidelines (ERG). These guidelines employ the use of Critical Safety Function Status Trees (CSFST).Since STP has implemented the WOG ERGs, the guidance in NEI 99-01 allows the use of certain CSFST assessment results as EALs and fission product barrier loss/potential loss thresholds.

This approach allows consistency between EOPs and emergency classifications. 711 P a e FUEL CLAD BARRIER THRESHOLDS The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.I1. RCS or SG Tube Leakage Loss I There is no Loss threshold associated with RCS or SG Tube LeakageL Potential Loss 1.A Core Cooling -Orange entry conditions (CETs > 7080 F) are sufficient to allow the onset of heat-induced cladding damage.2. Inadequate Heat Removal Loss 2.A Core Cooling -Red entry conditions (CETs > 12000 F) are sufficient to cause significant superheating of reactor coolant.Potential Loss 2.A Core Cooling -Orange entry conditions (CETs > 708' F) are sufficient to allow the onset of heat-induced cladding damage.Potential Loss 2.B Heat Sink -Red entry conditions met (NR level in all SG < 14% [34%] AND total AFW flow to SG <576 GPM). This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted. Meeting this threshold results in a SITE AREA EMERGENCY because this threshold is identical to RCS Barrier Potential Loss threshold 2.A; both will be met. This condition warrants a SITE AREA EMERGENCY declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and raise RCS pressure to the point where mass will be lost from the system.72 1 P a e FUEL CLAD BARRIER THRESHOLDS

3. RCS Activity / Containment Radiation Loss 3.A. 1 The readings for the containment high range area monitors (RT-8050 and RT-805 1) correspond to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300ýiCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage.Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. The values for RT-8050 and RT-8051 were based on Calculation STPNOC013-004 Rev.2. The threshold values were conservatively rounded within 2% of the calculated values to make the values readily assessable.

Temperature induced current (TIC) limitations are not applicable to the Fuel Clad Barrier Loss threshold 3.A. 1 because the expected radiation dose for this event overwhelms the TIC effect. This effect is discussed in the IOCFR50.59 evaluation 04-8245-60 associated with DCP 04-8245-3 3.Loss 3.A.2 The HATCH MONITOR is located outside containment and is the back-up monitor to the containment high range monitors (RT-8050 and RT-805 1). The HATCH MONITOR threshold value is based on Calculation No. 03-ZE-003. This value corresponds to the calculated containment high range monitor readings for Fuel Clad Barrier Loss 3.A The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold 3.A since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the EMERGENCY CLASSIFICATION LEVEL to a SITE AREA EMERGENCY. Loss 3.B This threshold indicates that RCS radioactivity concentration is greater than 300 [LCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.Potential Loss 3.There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.

4. Containment Integrity or Bypass Not Applicable (included for numbering consistency) 731 P a -e FUEL CLAD BARRIER THRESHOLDS

5. Other Indications Loss and/or Potential Loss 5.A N/A 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is lost.Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

741 P a o e RCS BARRIER THRESHOLDS The RCS Barrier includes the RCS primary side and its connections Lip to and including the pressurizer safety and relief valves, and other connections tip to and including the primary isolation valves.1. RCS or SG Tube Leakage Loss 1.A This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage.It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location -inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment. A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a SITE AREA EMERGENCY since the Containment Barrier Loss threshold I.A will also be met.Potential Loss I.A This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ECCS (SI)actuation has not occurred. The threshold is met when an operating procedure, or operating crew supervision, directs that a standby charging (makeup) pump be placed in service to restore and maintain pressurizer level.This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage.It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location -inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment. If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a SITE AREA EMERGENCY since the Containment Barrier Loss threshold L.A will also be met.Potential Loss I.B Integrity -Red entry conditions indicate an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock -a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized). 75 1 P a -e RCS BARRIER THRESHOLDS

2. Inadequate Heat Removal Loss 2.A There is no Loss threshold associated with Inadequate Heat Removal.Potential Loss 2.A Heat Sink- Red entry conditions met (NR level in all SG < 14% [34%] AND total AFW flow to SGs < 576 GPM).This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.

Meeting this threshold results in a SITE AREA EMERGENCY because this threshold is identical to Fuel Clad Barrier Potential Loss threshold 2.B; both will be met. This condition warrants a SITE AREA EMERGENCY declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and raise RCS pressure to the point where mass will be lost from the system.3. RCS Activity / Containment Radiation Loss 3.A.Not Applicable Potential Loss 3.There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.

4. Containment Integrity or Bypass Not Applicable (included for numbering consistency) 761 Pag e RCS BARRIER THRESHOLDS

5. Other Indications Loss and/or Potential Loss 5.A Variables used to monitor for the significant breach or the potential significant breach of fuel clad, the RCS pressure boundary, or the reactor Containment, are designated Type C. The response characteristics of Type C information display channels allow the control room operator to detect conditions indicative of significant failure of any of the three fission product barriers or the potential for significant failure of these barriers.

Although variables selected to fulfill Type C functions may rapidly approach the values that indicate an actual significant failure, it is the final steady-state value reached that is important. Therefore, a high degree of accuracy and a rapid response time are not necessary for Type C information display channels. Type C variables are found in UFSAR Table 7B.6- 1.6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in detennining whether the RCS Barrier is lost.Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. 77 1 P a e CONTAINMENT BARRIER THRESHOLDS Tile Containment Barrier includes the containment building and connections uip to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building uip to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the ECL from ALERT to a SITE AREA EMERGENCY or a GENERAL EMERGENCY.

l. RCS or SG Tube Leakage Loss L.A This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment.

The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss 1.A and Loss 1.A, respectively. This condition represents a bypass of the containment barrier.FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably [part of the FAULTED definition] and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU3 for the fuel clad barrier (i.e., RCS activity values) and IC SU4 for the RCS barrier (i.e., RCS leak rate values).This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment. Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do mneet this threshold. Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors., glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.781 Pa- e CONTAINMENT BARRIER THRESHOLDS The EMERGENCY CLASSIFICATION LEVELS resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.Affected SG is FAULTED Outside of Containment? P-to-S Leak Rate Yes No Less than or equal to 25 gpm Greater than 25 gpm Requires operation of a standby charging pump (RCS Barrier Potential Loss)Requires an automatic or manual ECCS (SI)actuation (RCS Barrier Loss)No classification UNUSUAL EVENT per SU4 SITE AREA EMERGENCY per FS I SITE AREA EMERGENCY per FS I No classification UNUSUAL EVENT per SU4 ALERT per FAl ALERT per FAI Potential Loss 1.There is no Potential Loss threshold associated with RCS or SG Tube Leakage.2. Inadequate Heat Removal Loss 2 There is no Loss threshold associated with Inadequate Heat Removal.Potential Loss 2.A Core Cooling -Red entry conditions met for 15 minutes or longer. This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and a higher potential for containment failure. For this condition to occur there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.The restoration procedure is considered "effective" if core exit thermocouple readings are decreasing and/or if RCS level is increasing. Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Director should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective. Severe accident analyses (e.g., NUREG-1 150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.791 P a g e CONTAINMENT BARRIER THRESHOLDS

3. RCS Activity / Containment Radiation Loss 3 There is no Loss threshold associated with RCS Activity / Containment Radiation.

Potential Loss 3.A.I The readings for the containment high range area monitors (RT-8050 and RT-805 1) correspond to an instantaneous release of the radioactive material inventory of the reactor coolant system (i.e., All the RCS coolant mass) into the containment, assuming that 20% of the fuel cladding has failed. The values for RT-8050 and RT-8051 were based on Calculation No. STPNOC013-004 Rev.2. The threshold values used were conservatively rounded within 2% of the calculated values to ensure the values were readily assessable. This level of assumed fuel clad failure is well beyond that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds. Temperature induced current (TIC) limitations are not applicable to the Containment Barrier Potential Loss threshold 3.A.I because the expected radiation dose for this event overwhelms the TIC effect. This effect is discussed in 1OCFR50.59 evaluation 04-8245-60 associated with DCP 04-8245-33. NUREG- 1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the EMERGENCY CLASSIFICATION LEVEL to a GENERAL EMERGENCY. Potential Loss 3.A.2 The HATCH MONITOR is located outside containment and is the back-up monitor to the containment high range monitors (RT-8050 and RT-805 1). The HATCH MONITOR threshold value is based on Calculation No. 03-ZE-003. This value corresponds to the calculated containment high range monitor readings for Containment Barrier Threshold Potential Loss 3.A. 1.4. Containment Integrity or Bypass Loss 4.A These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously mneet both thresholds 4.A.1 and 4.A.2.4.A.1 -Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage). Following the release of RCS mass into containment., containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate 80 1 P a , e CONTAINMENT BARRIER THRESHOLDS during accident conditions, it is expected that tile Emergency Director will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).Refer to the middle piping run of Figure 9-F-3. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside tile containment. Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.4.A.2 -Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g., through discharge of a ventilation system or atmospheric leakage).Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.Refer to the top piping run of Figure 9-F-3 in Addendum 3, Containment Integrity or Bypass Examples.In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the enviromnent. The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.Leakage between two interfacing liquid systems, by itself, does not meet this threshold. Refer to the bottom piping run of Figure 9-F-3. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the Component Cooling Water system to the Auxiliary Building, then no threshold has been met. If the pump or system piping developed a leak that allowed steam/water to enter the Auxiliary Building, then threshold 4.B would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A.1 to be met as well.81 Page CONTAINMENT BARRIER THRESHOLDS Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. Minor releases may also occur ifa containment isolation valve(s)fails to close but the containment atmosphere escapes to a closed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold I.A.Loss 4.B Containment sump, temperature, pressure and/or radiation levels will rise if reactor coolant mass is leaking into the containment. If these parameters have not risen, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). Rises in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment. Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not rise significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc. should be sufficient to determine if RCS mass is being lost outside of the containment. Refer to the middle piping run of Figure 9-F-3. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A.I to be met as well.To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold I.A to be met.Potential Loss 4.A Containment -Red entry conditions met (containment pressure > 56.5 PSIG). If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a SITE AREA EMERGENCY and GENERAL EMERGENCY since there is now a potential to lose the third barrier.Potential Loss 4.B The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen bum (i.e., at the lower deflagration limit (4%)). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a potential loss of the Containment Barrier.82 1 P a g e CONTAINMENT BARRIER THRESHOLDS Potential Loss 4.C This threshold describes a condition where containment pressure is greater than the setpoint (9.5 PSIG) at which Contaimnent Spray is designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that Containment Spray is either lost or performing in a degraded manner.5. Other Indications Loss and/or Potential Loss 5.A N/A 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost.Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. 83 I P a -e Figure 9-F-3: Containment Integrity or Bypass Examples RCP Seal Cooling NOTES: Only Supplemental Purge is a filtered release and STPEGS Component Cooling Water is equivalent to Closed Cooling Water 841 P ag e 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS Table H-I: Recognition Category "H" Initiating Condition Matrix UNUSUAL EVENT HU1 Confirmed SECURITY CONDITION or threat.Op. Modes: ALL ALERT SITE AREA EMERGENCY GENERAL EMERGENCY HU2 Seismic event greater than OBE levels.Op. Modes: ALL HU3 Hazardous event.Op. Modes: ALL HU4 FIRE potentially degrading the level of safety of the plant.Op. Modes: ALL HA1 HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.Op. Modes: ALL Note: See SA9 or CA6 for escalation of these events HA5 Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.Op. Modes: ALL HA6 Control Room evacuation resulting in transfer of plant control to alternate locations. Op.AModes: ALL HA7 Other conditions exist which in the judgment of the Emergency Director warrant declaration of an ALERT.Op. Modes: ALL HS1 HOSTILE ACTION within the PROTECTED AREA.Op. Modes: ALL HG1 HOSTILE ACTION resulting in loss of physical control of the facility.Op. Modes: ALL HU7 Other conditions exist which in the judgment of the Emergency Director warrant declaration of an UNUSUAL EVENT.Op. Modes: ALL HS6 Inability to control a key safety function from outside the Control Room.Op. Modes: ALL HS7 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY. Op. Alodes: ALL HG7 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY. Op. AMlodes: ALL 85 I P a- e HU1 ECL: UNUSUAL EVENT Initiating Condition: Confirmed SECURITY CONDITION or threat.Operating Mode Applicability: ALL Emergency Action Levels: (1 or 2 or 3)(1) A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by ANY of the following personnel in Table HI: Table Hi: Security Supervision

• Security Force Supervisor
• Acting Security Manager* Security Manager (2) Notification of a CREDIBLE SECURITY THREAT directed at the site.(3) A validated notification from the NRC providing information of an aircraft threat.Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. SECURITY EVENTS which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72..SECURITY EVENTS assessed as HOSTILE ACTIONS are classifiable tinder ICs HAl, HSI and HG I.Timely and accurate communications between Security Force Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and OROs.Security plans and terminology are based on the guidance provided by NEI 03-12, Template./br the Security Plan, Training and Qualification Plan, Sa feguards Contingency Plan [and INDEPENDENT SPENT FUEL STORAGE INSTALLATION Security Program].EAL #1- references Security Force Supervisor because these are the individuals trained to confirm that a SECURITY EVENT is occurring or has occurred.

Training on SECURITY EVENT confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39039 information. EAL #2- addresses the receipt of a CREDIBLE SECURITY THREAT. The credibility of the threat is assessed in accordance with OSDPOI-ZS-0011, Implementing Procedure For Safeguards Contingency Events.EAL #3- addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is 861Pag e performed in accordance with OPOP04-ZO-SEC4. Guideline For Airborne (Aircraft) Threat, and Security Force Instruction SI 2700, Security Response to Airborne Threat.Emergency plans and implementing procedures are public documents; therefore, EALs do not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information is contained in the Security Plan.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC HAl.HUI: EAL-1 Selection Basis: For EAL-1, the position of Security Force Supervisor was included since it is a 24-hour position.Normally the event would not be reported by the Acting Security Manager or Security Manager because the Acting Security Manager position is not normally activated until after an UNUSUAL EVENT has been declared, and the Security Manager position is not normally activated until after an ALERT has been declared. However, reporting by the Acting Security Manager or Security Manager was included in the event these positions are staffed under unusual circumstances. REFERENCEs:

1. OERPOI-ZV-SH03, Rev. 12, Acting Security Manager 2. OERPO1-ZV-TS08, Rev. 16, Security Manager 3. 0POP04-ZO-SEC4, Rev. 10, Guideline For Airborne (Aircraft)

Threat (SUNS1)4. OSDPO1-ZS-001 1, Implementing Procedure For Safeguards Contingency Events (Safeguards)

5. Security Force Instruction SI 2700, Security Response to Airborne Threat (SUNSI)871 a t, e HU2 ECL: UNUSUAL EVENT Initiating Condition:

Seismic event greater than OBE levels.Operating Mode Applicability: ALL Emergency Action Level: (1) a. EITHER of the following conditions exist: 1. "SEISMIC EVENT" alarm in Unit I Control Room (Lampbox 9M0I, Window E-8)OR 2. Control Room personnel feel an actual or potential seismic event.AND b. The occurrence of a seismic event is confirmed in manner deemed appropriate by the Shift Manager or Emergency Director.Basis: This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.Although the "SEISMIC EVENT" alarm (0.02 g) in EAL L.a is set below an O.B.E earthquake (0.05 g), it does provide an indication that a seismic event has occurred. In order to determine whether an O.B.E.earthquake occurred, additional indications may be needed. Determination per 0POP04-SY-001, Seismic Event is not practical if it takes longer than 15 minutes to perform.Indications described in the EAL should be limited to those that are immediately available to Control Room personnel and which can be readily assessed. Indications available outside the Control Room and/or which require lengthy times to assess (e.g., processing of scratch plates or recorded data) should not be used. The goal is to specify indications that can be assessed within 15-minutes of the actual or suspected seismic event.The EAL 1.b- statement is included to ensure that a declaration does not result from felt vibrations caused by a non-seismic source (e.g., a dropped heavy load). The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration. It is recognized that this alternate EAL wording may cause a site to declare an UNUSUAL EVENT while another site, similarly affected but with readily assessable OBE indications in the Control Room, may not.88 1 P a o e Depending upon the plant mode at the time of the event, escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC CA6 or SA9.HU2: EAL-1 Selection Basis: STP does not have a readily available indication in the Control Room for determining if the site has experienced an OBE. The Seismic Event Alarm setpoint is 0.02g in the vertical or horizontal position and the station design basis value for an OBE is 0.05g. Since the Seismic Event alarm is set at less than half of the OBE value, it cannot be used as the sole threshold value for detennining whether or not STP has experienced an OBE.STP has implemented the alternative EAL described in NEI 99-01 Developer Notes in conjunction with using the installed indication. EAL-1, b. allows the Shift Manager or Emergency Director to determine if a seismic event has taken place, taking into consideration the Seismic Event alarm, Control Room personnel feeling an actual or potential seismic event and other indications deemed appropriate.

REFERENCES:

1. OPOP04-SY-0001, Rev. 8, Seismic Event 2. NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors.89 1 P a ge HU3 ECL: UNUSUAL EVENT Initiating Condition:

Hazardous event.Operating Mode Applicability: ALL Emergency Action Levels: (1 or 2 or 3 or 4 or 5)Note: EAL #4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. (1) A tornado strike within the PROTECTED AREA.(2) Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode.(3) Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).(4) A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.(5) Predicted or actual breach of Main Cooling Reservoir retaining dike along North Wall Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.EAL #1 -addresses a tornado striking (touching down) within the PROTECTED AREA.EAL #2- addresses flooding of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.EAL #3- addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.EAL #4- addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff fi'om accessing the site using personal vehicles. Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road. This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around 90 I P a -e the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.EAL#5- the Main Cooling Reservoir breach along the north wall which was included because it is a credible hazard and analyzed in the STPEGS UFSAR.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be based on ICs in Recognition Categories R, F, S or C.HU3: EAL-1, EAL-2, EAL-3, EAL-4 Selection Basis: N/A

REFERENCE:

1. STPEGS UFSAR, Section 3.4.1, Flood Protection 911 1 a " e HU4 ECL: UNUSUAL EVENT Initiating Condition:

FIRE potentially degrading the level of safety of the plant.Operating Mode Applicability: ALL Emergency Action Levels: (1 or 2 or 3 or 4)Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.(1) a. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications:

• Report from the field (i.e., visual observation)
• Receipt of multiple (more than 1) fire alarms or indications
• Field verification of a single fire alarm AND b. The FIRE is located within ANY of the plant rooms or areas in Table H4: Table H4: Plant Rooms/Areas
• Mechanical/Electrical Auxiliary Building (MEAB)* Fuel Handling Building (FHB)* Reactor Containment Building (RCB)* Essential Cooling Water Intake Structure (ECWIS)* Isolation Valve Cubicle (IVC)* Diesel Generator Building (DGB)(2) a. Receipt of a single fire alarm (i.e., no other indications of a FIRE).AND b. The FIRE is located within ANY of the plant rooms or areas in Table H4: AND c. The existence of a FIRE is not verified within 30-minutes of alarm receipt.(3) A FIRE within the ISFSI OR plant PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication.

(4) A FIRE within the ISFS1 OR plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish. 921 P age Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.EAL #1 The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.EAL #2 This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed. A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.If an actual FIRE is verified by a report from the field, then EAL #1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. EAL #3 In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant or ISFSI PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety.EAL #4 If a FIRE within the plant or ISFSI PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.93 1P a g e Basis-Related Reouirements from Annendix R Appendix R to 10 CFR 50, states in part: Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and EXPLOSIONS." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in EAL #2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.Depending upon the plant mode at the time of the event, escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC CA6 or SA9.HU4: EAL-1.b. EAL-2.b Selection Basis: The plant areas or rooms listed contain SAFETY SYSTEM equipment.

REFERENCES:

1. OPGP03-ZF-0001, Rev. 26, Fire Protection Program 2. STPEGS UFSAR, Rev. 16, Section 7.4, Systems Required for Safe Shutdown 94 P a g e HU7 ECL: UNUSUAL EVENT Initiating Condition:

Other conditions exist which in the judgment of the Emergency Director warrant declaration of a UE.Operating Mode Applicability: ALL Emergency Action Level: (1) Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to FACILITY protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the EMERGENCY CLASSIFICATION LEVEL description for an UE.HU7: EAL-1 Selection Basis: N/A

REFERENCES:

N/A 95 1 P a o e HA1 ECL: ALERT Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.Operating Mode Applicability: ALL Emergency Action Levels: (1 or 2)(1) A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by ANY of the following personnel in Table HI: Table HI: Security Supervision

• Security Force Supervisor
• Acting Security Manager" Security Manager (2) A validated notification from NRC of an aircraft attack threat within 30 minutes of the site.Basis: This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Securitv Plan. Training and Qualification Plan, Safeguards Contingency Plan [and INDEPENDENT SPENT FUEL STORAGE INSTALLA TION Security Program].As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering).

The ALERT declaration will also heighten the awareness of Offsite Response Organizations, allowing them to be better prepared should it be necessary to consider further actions.This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs., or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.EAL 91- is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA.961Pagc EAL #2 addresses the threat friom tile impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with OPOP04-ZO-SEC4, Guidelines for Airborne (Aircraft) Threat, and Security Force Instruction SI 2700, Security Response to Airborne Threat.The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to tile site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency.Emergency plans and implementing procedures are public documents; therefore, EALs do not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive infonnation is contained in the Security Plan.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC HSI.HAl: EAL-1 and EAL-2 Selection Basis: The EALs are taken from NEI 99-01, Rev. 6. For EAL-l, the positions of Security Force Supervisor OR Acting Security Manager were included because either of these positions could be activated prior to meeting this EAL. The Security Force Supervisor is a 24-hour position and the normally the Acting Security Manager is activated after an UNUSUAL EVENT has been declared. The Security Manager is also included although this position is normally activated after an ALERT.

REFERENCES:

1. OERPO1-ZV-SH03, Rev. 12, Acting Security Manager 2. OERPO1-ZV-TS08, Rev. 16, Security Manager 3. OPOP04-ZO-SEC4, Rev. 10, Guideline For Airborne (Aircraft)

Threat (SUNSI)4. Security Force Instruction SI 2700, Security Response to Airborne Threat (SUNSI)97 I P a e HA5 ECL: ALERT Initiating Condition: Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.Operating Mode Applicability: ALL Emergency Action Level: Note: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. (1) a. Release of a toxic, corrosive, asphyxiant or flammable gas into the Control Room or ANY of the plant rooms or areas listed in Table H3/R2: AND b. Entry into the room or area is prohibited or impeded.TABLE H3/R2: Plant Areas Requiring Access RCB RHR Heat Exchanger Rooms 0 C ) MAB 51 ft Room 335 EAB Roof, MCC 1G8, 4.16KV Switchgear Rooms 0 0 Ln EAB 4.16KV Switchgear Rooms 98 1P a , e Basis: This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant.An ALERT declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release.Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). An emergency declaration is not warranted if any of the following conditions apply." The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release).* For example, the plant is in Mode I when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4." The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).* The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

• The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment.

This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.This EAL does not apply to firefighting activities that automatically or manually activate a fire suppression system in an area.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via Recognition Category R, C or F ICs.99 1 P a g e HA5: EAL-1 Selection Basis: The areas listed in EAL-1 apply to areas that contain equipment necessary for plant operations, cooldown, or shutdown. Assuming all plant equipment is operating as designed., Normal operations and safe shutdown equipment operation is capable from the Main Control Room (MCR). The plant is able to transition into a hot shutdown from the MCR, therefore H3/R2 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary following entry into hot shutdown (establish Residual Heat Removal shutdown cooling, disable operation of charging and ECCS equipment, and limit dilution pathways) and subsequent entry into cold shutdown (disable operation of ECCS equipment). After achieving cold shutdown it is assumed that the plant will be maintained in a cold shutdown condition.

REFERENCES:

1. OPGP03-ZF-0001, Rev. 26, Fire Protection Program 2. STPEGS UFSAR, Rev. 16, Section 7.4, Systems Required for Safe Shutdown 3. 0POP03-ZG-0008, Rev. 56, Power Operations

4. OPOP03-ZG-0006, Rev. 54, Plant Shutdown from 100% to Hot Standby 5. OPOP03-ZG-0007, Rev. 71, Plant Cooldown 100 P ag e HA6 ECL: ALERT Initiating Condition:

Control Room evacuation resulting in transfer of plant control to alternate locations. Operating Mode Applicability: ALL Emergency Action Level: (1) An event has resulted in plant control being transferred firom the Control Room to the Auxiliary Shutdown Panel (ASP).Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift persormel. Activation of the ERO and emergency response facilities will assist in responding to these challenges. Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC HS6.HA6: EAL-1 Selection Basis: The Auxiliary Shutdown Panel (ASP) is identified in OPOP04-ZO-0001, Control Room Evacuation, as the location where plant control is transferred in the event of a Control Room evacuation.

REFERENCES:

I. Procedure OPOP04-ZO-0001, Rev. 35, Control Room Evacuation 101 lPage HA7 ECL: ALERT Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of an ALERT.Operating Mode Applicability: ALL Emergency Action Level: (1) Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a SECURITY EVENT that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. ANY releases are expected to be limited to small fractions of the EPA PROTECTIVE ACTION GUIDELINE exposure levels.Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the EMERGENCY CLASSIFICATION LEVEL description for an ALERT.HA7: EAL-1 Selection Basis: N/A

REFERENCE:

N/A 102 1 P age HS1 ECL: SITE AREA EMERGENCY Initiating Condition: HOSTILE ACTION within tile PROTECTED AREA.Operating Mode Applicability: ALL Emergency Action Level: (1) A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by ANY of the following personnel in Table HI: Table Hi: Security Supervision

• Security Force Supervisor
• Acting Security Manager* Security Manager Basis: This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Securitv Plan, Training and Qualification Plan, Sqfeguards Contingency Plan [and INDEPENDENT SPENT FUEL STORAGE INSTALL4 TION Security Program].As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The SITE AREA EMERGENCY declaration will mobilize ORO resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions. This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.Emergency plans and implementing procedures are public documents; therefore, EALs do not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive infonnation is contained in the Security Plan.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC HGI.1031 P a c HSI: EAL-I Selection Basis: The positions of Security Force Supervisor, Acting Security Manager, and Security Manager were included since any of these positions could be activated prior to meeting this EAL. The Security Force Supervisor is a 24-hour position, the Acting Security Manager is activated after an Unusual Event has been declared and the Security Manager is activated after an Alert is declared.

REFERENCES:

I. OERPOI-ZV-SH03, Rev. 12, Acting Security Manager 2. OERPO1-ZV-TS08, Rev. 16, Security Manager 104 P ag, e HS6 ECL: SITE AREA EMERGENCY Initiating Condition: Inability to control a key safety function from outside the Control Room.Operating Mode Applicability: ALL Emergency Action Level: Note: The Emergency Director should declare the SITE AREA EMERGENCY promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panel (ASP).AND b. Control of ANY of the following key safety functions in Table H2 is not reestablished within 15 minutes in Modes 1, 2 or 3 ONLY.Table H2: Safety Functions* Reactivity control* Core cooling* RCS heat removal Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manmer. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.The determination of whether or not "control" is established at the Auxiliary Shutdown Panel is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s). Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC FGI or CGI.HS6: EAL-1 Selection Basis: The Auxiliary Shutdown Panel (ASP) is identified in OPOP04-ZO-0001, Control Room Evacuation, as the location where plant control is transferred in the event of a Control Room evacuation. The 15 minute timeframe to control the key safety functions is identified as site specific information. The mode applicability conditioning statement for Table H2 is based on the Technical Specification Operability requirement for the following functions of the Remote Shutdown System:* Core reactivity control (initial and long term)105 1 P a g e

• RCS pressure control" Decay heat removal via the AFW System and the SG safety valves or SG PORVs" RCS inventory control via charging flow, and* Safety support systems for the above functions.

REFERENCE:

I. Procedure OPOP04-ZO-0001, Rev. 35, Control Room Evacuation

2. Technical Specification 3.3.3.5 Remote Shutdown System 106 1 P a o e HS7 ECL: SITE AREA EMERGENCY Initiating Condition:

Other conditions exist which in the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY. Operating Mode Applicability: ALL Emergency Action Level: (1) Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. ANY releases are not expected to result in exposure levels which exceed EPA PROTECTIVE ACTION GUIDELINE exposure levels beyond the SITE BOUNDARY.Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the EMERGENCY CLASSIFICATION LEVEL description for a SITE AREA EMERGENCY. HS7: EAL-1 Selection Basis: N/A

REFERENCE:

N/A 107 1 P a -e HG1 ECL: GENERAL EMERGENCY Initiating Condition: HOSTILE ACTION resulting in loss of physical control of the FACILITY.Operating Mode Applicability: ALL Emergency Action Level: (1) a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by ANY of the following in Table HI1: Table HI: Security Supervision

• Security Force Supervisor
• Acting Security Manager* Security Manager AND b. EITHER of the following has occurred: 1. ANY of the following safety functions in Table H2 cannot be controlled or maintained in MODES 1,2 or 3 ONLY.Table H2: Safety Functions* Reactivity control" Core cooling* RCS heat removal OR 2. Damage to spent fuel has occurred or is IMMINENT.Basis: This IC addresses an event in which a HOSTILE FORCE has taken physical control of the FACILITY to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions.

It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system (e.g., pumps, heat exchangers., controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Securitv Plan, Training and Qualification Plan, Safeguards Contingency Plan [and INDEPENDENT SPENT FUEL STOR4GE INSTALLA TION Securiz, Program].108 1 P a o e Emergency plans and implementing procedures are public documents; therefore, EALs do not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information is contained in the Security Plan.HG1: EAL-1 Selection Basis: The positions of Security Force Supervisor, Acting Security Manager, and Security Manager were also included since any of these positions could be activated prior to meeting this EAL. The mode applicability conditioning statement for Table H2 is based on the Technical Specification Operability requirement for the following Functions of the Remote Shutdown System:* Core reactivity control (initial and long term)" RCS pressure control* Decay heat removal via the AFW System and the SG safety valves or SG PORVs* RCS inventory control via charging flow, and* Safety support systems for the above Functions.

REFERENCES:

1. OERPO1-ZV-SH03, Rev. 12, Acting Security Manager 2. OERPOI-ZV-TS08, Rev. 16, Security Manager 3. Technical Specification 3.3.3.5 Remote Shutdown System 109 P a -e HG7 ECL: GENERAL EMERGENCY Initiating Condition:

Other conditions exist which in the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY. Operating Mode Applicability: ALL Emergency Action Level: (1) Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the FACILITY. Releases call be reasonably expected to exceed EPA PROTECTIVE ACTION GUIDELINE exposure levels offsite for more than the immediate site area.Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall tinder the EMERGENCY CLASSIFICATION LEVEL description for a GENERAL EMERGENCY. HG7: EAL-1 Selection Basis: N/A

REFERENCE:

N/A 110 1pa e 11 SYSTEM MALFUNCTION ICS/EALS Table S-1: Recognition Category "S" Initiating Condition Matrix UNUSUAL EVENT ALERT SITE AREA EMERGENCY SU1 Loss of ALL offsite AC power capability to emergency buses for 15 minutes or longer.Op. Modes: 1,2,3,4 SU2 UNPLANNED loss of Control Room indications for 15 minutes or longer.Op. Modes: 1,2,3,4 SU3 Reactor coolant activity greater than Technical Specification allowable limits.Op. Modes: 1,2,3,4 SU4 RCS leakage for 15 minutes or longer.Op. Modes: 1,2,3,4 SU5 Automatic or manual trip fails to shutdown the reactor.Op. Modes: 1,2 SA1 Loss of ALL but one AC power source to emergency buses for 15 minutes or longer.Op. Modes: 1,2,3,4 SA2 UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.Op. Modes: 1,2,3.4 SS1 Loss of ALL offsite and ALL onsite AC power to emergency buses for 15 minutes or longer.Op. Modes: 1,2,3,4 GENERAL EMERGENCY SG1 Prolonged loss of ALL offsite and ALL onsite AC power to emergency buses.Op. Modes.: 1,2,3,4 SA5 Automatic or manual trip fails to shutdown the reactor, and subsequent manual actions taken at the reactor control panels are not successful in shutting down the reactor.Op. Alodes: 1,2 SS5 Inability to shutdown the reactor causing a challenge to core cooling or RCS heat removal.Op. Af-odes: 1,2 III IP a" e Table S-1: Recognition Category "S" Initiating Condition Matrix (cont.)UNUSUAL EVENT ALERT SITE AREA GENERAL EMERGENCY EMERGENCY SU6 Loss of ALL onsite or offsite communications capabilities. Op. Modes: 1,2,3.4 SU7 Failure to isolate containment or loss of containment pressure control. 1,2,3,4 SS8 Loss of ALL Vital DC SG8 Loss of ALL AC and power for 15 minutes or Vital DC power sources for longer. 15 minutes or longer. Op.Op. Modes: 1,2,3,4 MIodes: 1,2,3,4 SA9 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.Op. Modes: 1,2,3,4 112 P age Sul ECL: UNUSUAL EVENT Initiating Condition: Loss of ALL offsite AC power capability to emergency buses for 15 minutes or longer.Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Level: Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Loss of ALL offsite AC power capability to ALL three 4160V AC ESF Buses for 15 minutes or longer.Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency buses. This condition represents a potential reduction in the level of safety of the plant.For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC SA1.SUl: EAL-1 Selection Basis: N/A

REFERENCES:

I. OPOP04-AE-0001, Rev. 44, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus 2. OPOP04-AE-0004, Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus 3. 0PSP03-EA-0002, Rev. 32, ESF Power Availability

4. Drawing 00000EOAAAA, Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. 1&2 113 1 P a o e SU2 ECL: UNUSUAL EVENT Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer.Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Level: Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) An UNPLANNED event results in the inability to monitor one or more of the following parameters in Table S I from within the Control Room for 15 minutes or longer.Table Si: Plant Parameters

• Reactor Power* RCS Level* RCS Pressure* Core Exit Temperature
• Levels in at least two steam generators
• Steam Generator Auxiliary Feed Water Flow Basis: This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s).

For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantlyimpaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, 114 P a o e then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well.For example, if the value for RCS level cannot be determined from tile indications and recorders on a main control board, the SPDS or the plant computer, tile availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC SA2.SU2: EAL-1 Selection Basis: The parameters listed were from NEI 99-0 1, Rev. 6 with the exception of steam generators. Two steam generators is a site-specific parameter for the minimum number of steam generators needed for plant cooldown and shutdown.

REFERENCES:

1. OPOP05-EO-E020, Rev. 11, Faulted Steam Generator Isolation 2. OPOP05-EO-FRHI, Rev. 23, Response to Loss of Secondary Heat Sink 115 1 P a g e SU3 ECL: UNUSUAL EVENT Initiating Condition:

Reactor coolant activity greater than Technical Specification allowable limits.Operating Mode Applicability: 1,2, 3, 4 Emergency Action Levels: (1 or 2)(1) RT-8039 reading greater than 30 pCi/cm 3.(2) Sample analysis indicates that a reactor coolant activity value is greater than an allowable limit specified in Technical Specifications.

• Greater than I laCi/gm Dose Equivalent 1-131* Greater than 100/ E bar [iCi /gn gross activity Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications.

This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via ICs FAI or the Recognition Category R ICs.SU3: EAL-1 Selection Basis: RT-8039 is the Failed Fuel radiation monitor and samples via the CVCS letdown line. The value 30 pCi/cm 3 is the reading that is equivalent to I pCi/gm Dose Equivalent 1-131. The monitor value in this EAL is the calculated monitor response if the RCS activity were equivalent to 1 [Ci/gm Dose Equivalent 1-131. The value is based on Calculation STPNOC013-CALC-003. The value used in this EAL was conservatively truncated by approximately 5% to ensure the value is readily assessable. SU3: EAL-2 Selection Basis: The Technical Specification limits for RCS activity is greater than I pCi/gm Dose Equivalent 1-131 or greater than I 00/F bar puCi /gn gross activity.

REFERENCES:

1. Calculation No. STPNOC013-CALC-003 Rev.1, Gross Failed Fuel Monitor Response to Rise RCS Activity (RT-8039 EAL Threshold)

2. STP Technical Specification Section 3/4.4.8 Specific Activity.116 1P a ge SU4 ECL: UNUSUAL EVENT Initiating Condition:

RCS leakage for 15 minutes or longer.Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels: (1 or 2 or 3)Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) RCS unidentified or pressure boundary leakage greater than 10 gpm for 15 minutes or longer.(2) RCS identified leakage greater than 25 gpm for 15 minutes or longer.(3) Leakage firom the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer.Basis: This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.EAL #1 and EAL #2 are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). EAL #3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage) or a location outside of containment. The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). EAL #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated). The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via ICs of Recognition Category RorF.117 1 P a e SU4: EAL-1 Selection Basis: The STP Technical Specifications limit for unidentified leakage from the RCS is 1 gpm. NEI 99-01 Rev.6 states to use the higher of the Technical Specification limit or 10 gpm.SU4: EAL-2 Selection Basis: Tile STP Technical Specifications limit for identified leakage from the RCS is 10 gpm. NEI 99-01 Rev. 6 requirements are to use the higher of the Technical Specification limit or 25 gpm.SU4: EAL-3 Selection Basis: The STP Technical Specification limit for primary-to-secondary leakage is 150 gallons per day through any one steam generator, but the specification does not specify the type of leakage. Therefore, STPEGS will use the leakage outside containment; which may include SG Tube Leakage, at 25 gpm for 15 minutes or longer in accordance with NEI 99-01 Rev. 6 guidance.EFERENCES:

1. STP Technical Specification Section 3.4.6.2 Reactor Coolant System Operational Leakage.118 P a o e SU5 ECL: UNUSUAL EVENT Initiating Condition:

Automatic or manual trip fails to shutdown the reactor.Operating Mode Applicability: 1, 2 Emergency Action Levels: (I or 2)Note: A manual action is ANY operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. (1) a. An automatic trip did not shutdown the reactor.AND b. A subsequent manual action taken at the reactor control panels is successful in shutting down the reactor.(2) a. A manual trip did not shutdown the reactor.AND b. EITHER of the following: I. A subsequent manual action taken at the reactor control panels is successful in shutting down the reactor.OR 2. A subsequent automatic trip is successful in shutting down the reactor.Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control panels or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control panels to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.If an initial manual reactor trip is unsuccessfiul, operators will promptly take manual action at another location(s) on the reactor control panels to shut down the reactor (e.g., initiate a manual reactor trip) using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a 119 1 Pag e subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.A manual action at the reactor control panels is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual trip). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control panels".The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control panels are also unsuccessful in shutting down the reactor, then the EMERGENCY CLASSIFICATION LEVEL will escalate to an ALERT via IC SA5. Depending upon the plant response, escalation is also possible via IC FA 1. Absent the plant conditions needed to meet either IC SA5 or FA 1, an UNUSUAL EVENT declaration is appropriate for this event.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.Should a reactor trip signal be generated as a result of plant work (e.g.. RPS setpoint testing), the following classification guidance should be applied.* If the signal causes a plant transient that should have included an automatic trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

• If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.

SU5: EAL-1 EAL-2 Selection Basis: N/A

REFERENCES:

I. OPOP03-ZG-0004, Rev. 45, Reactor Startup 2. 0POP03-ZG-0005, Rev. 86, Plant Startup to 100%120 1 Page SU6 ECL: UNUSUAL EVENT Initiating Condition: Loss of ALL onsite or offsite communications capabilities. Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels: (1 or 2 or 3)(1) Loss of ALL of the following onsite communication methods listed in Table S2.(2) Loss of ALL of the following Offsite Response Organization (ORO) communications methods listed in Table S2.(3) Loss of ALL of the following NRC communications methods listed in Table S2.Table S2: Communications Methods EAL-1 EAL-2 EAL-3 ONSITE ORO NRC* Plant PA system X* Plant Radios X" Plant telephone system X X X" Satellite phones X X* Direct line from Control Rooms to Bay X X City" Microwave Lines to Houston X X* Security radio to Matagorda County X* Dedicated Ring-down lines X* ENS line X Basis: This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).//EAL #1- addresses a total loss of the commLunications methods used in support of routine plant operations. 121 IPaoe EAL #2- addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are Matagorda County Sheriff s Office, and Texas Department of Public Safety Disaster District in Pierce.EAL #3- addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. SU6: EAL-1, EAL-2, EAL-3 Selection Basis: Lines not included for offsite communications to ORO and NRC included links that would need relaying of information. Links were obtained from procedures OPGP05-ZV-0011, Emergency Communications.

REFERENCES:

1. OPGP05-ZV-001 1, Emergency Communications 122 1 P a g e SU7 ECL: UNUSUAL EVENT Initiating Condition:

Failure to isolate containment or loss of containment pressure control.Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels: (1 or 2)(1) a. Failure of containment to isolate when required by an actuation signal.AND b. ALL required penetrations are not isolated within 15 minutes of the actuation signal.(2) a. Containment pressure greater than 9.5 psig.AND b. No Containment Spray train is operating per design for 15 minutes or longer.Basis: This IC addresses a failure of one or more containment penetrations to automatically isolate when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.EAL #1- the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status -isolated or not isolated -should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.EAL #2- addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-1minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment spray) are either lost or performing in a degraded manner.This event would escalate to a SITE AREA EMERGENCY in accordance with IC FSI if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.1231 Paoe SU7: EAL-1 Selection Basis: N/A SU7: EAL-2 Selection Basis: If containment pressure reaches 9.5 psig, Containment Spray will actuate. If no train of Containment Spray is operating per design, the ability to lower containment pressure is compromised. One train of Containment Spray (Technical Specifications 3/4.6.2) is defined as one containment spray system capable of taking a suction from the RWST and transferring suction to the containment sump.

REFERENCES:

1. OPOP05-EO-F005, Rev. 1, Containment Critical Safety Function Status Tree 2. OPOP05-EO-FRZ1, Rev. 9, Response to High Containment Pressure 3. Technical Specifications 3/4.6.2 1241 P a o e SA1 ECL: ALERT Initiating Condition:

Loss of ALL but one AC power source to emergency buses for 15 minutes or longer.Operating Mode Applicability: 1, 2. 3, 4 Emergency Action Level: Note: The Emergency Director should declare the ALERT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. AC power capability to ALL three 4160V AC ESF Buses is reduced to a single power source for 15 minutes or longer.AND b. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU 1.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

• A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from the unit main generator.
• A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed firom an onsite or offsite power source.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC SS1.SAt: EAL-1 Selection Basis: This EAL is similar to IC CU2, except this EAL applies only to Modes 1-4.125 P11 a e

REFERENCES:

I. OPOP04-AE-0001, Rev. 44, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus 2. OPOP04-AE-0004, Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus 3. OPSP03-EA-0002, Rev. 32, ESF Power Availability

4. Drawing 00000EOAAAA, Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. I 126 Pa e c SA2 ECL: ALERT Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Level: Note: The Emergency Director should declare the ALERT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. An UNPLANNED event results in tile inability to monitor one or more of the following parameters in Table S I from within the Control Room for 15 minutes or longer.Table Si: Plant Parameters

• Reactor Power* RCS Level* RCS Pressure* Core Exit Temperature
• Levels in at least two steam generators
• Steam Generator Auxiliary Feed Water Flow AND b. ANY of the following transient events in progress.* Automatic or manual runback greater than 25% thermal reactor power* Electrical load rejection greater than 25% full electrical load* Reactor trip" ECCS (SI) actuation Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room.During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s).

For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.127 1P a () e An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well.For example, if the value for RCS level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via ICs FSI or IC RSI.SA2: EAL-1 Selection Criteria: The plant parameters listed are from NEI 99-01, Rev. 6. Two steam generators were selected as a site-specific parameter for the minimum number of steam generators needed for plant cooldown and shutdown.

REFERENCES:

I. OPOP05-EO-EO20, Rev. 11, Faulted Steam Generator Isolation 2. OPOP05-EO-FRHI1, Rev. 23, Response to Loss of Secondary Heat Sink 1281 P a ge SA5 ECL: ALERT Initiating Condition: Automatic or manual trip fails to shutdown the reactor, and subsequent manual actions taken at the reactor control panels are not successful in shutting down the reactor.Operating Mode Applicability: 1, 2 Emergency Action Level: Note: A manual action is ANY operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. (1) a. An automatic or manual trip did not shutdown the reactor.AND b. Manual actions taken at the reactor control panels are not successful in shutting down the reactor.Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control panels to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control panels since this event entails a significant failure of the RPS.A manual action at the reactor control panels is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s)is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control panels (e.g., locally opening breakers). Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control panels".The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the EMERGENCY CLASSIFICATION LEVEL will escalate to a SITE AREA EMERGENCY via IC SS5. Depending upon plant responses and symptoms, escalation is also possible via IC FSI.It is recognized that plant responses or symptoms may also require an ALERT declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration. 129 1 P age A reactor shutdown is detenmined in accordance with applicable Emergency Operating Procedure criteria.SA5: EAL-1 Selection Basis: N/A

REFERENCES:

1. OPOP05-EO-FRSl, Rev. 17, Response to Nuclear Power Generation

-ATWS 130 Paoe SA9 ECL: ALERT Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Level: (1) a. The occurrence of ANY of the following hazardous events listed in Table S3: Table S3: Hazardous Events* Seismic event (earthquake)" hIternal or external flooding event* High winds or tornado strike" FIRE* EXPLOSION* Predicted or actual breach of Main Cooling Reservoir retaining dike along North Wall.* Other events with similar hazard characteristics as determined by the Shift Manager AND b. EITHER of the following:

1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.OR 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.Basis: This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.EAL# 1.b. 1- addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.

The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.EAL# I.b.2- addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. 1311 Pa e Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC FSI or RSI.SA9: EAL-1 Selection Basis: The listed hazards are from NEI 99-01 Rev.6 with the exception of the Main Cooling Reservoir breach along the north wall which was included because it is a credible hazard and analyzed in the STPEGS UFSAR.

REFERENCES:

1. STPEGS UFSAR, Section 3.4.1, Flood Protection 132 1P a g e SS1 ECL: SITE AREA EMERGENCY Initiating Condition:

Loss of ALL offsite and ALL onsite AC power to emergency buses for 15 minutes or longer.Operating Mode Applicability: 1, 2, 3. 4 Emergency Action Level: Note: The Emergency Director should declare the SITE AREA EMERGENCY promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Loss of ALL offsite AND ALL onsite AC power to ALL tlhree 4160V AC ESF Buses for 15 minutes or longer.Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via ICs RGI, FGI or SGI.SS1: EAL-1 Selection Criteria: N/A

REFERENCES:

I. OPOP04-AE-0001, Rev. 44, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus 2. OPOP04-AE-0004, Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus 3. OPSP03-EA-0002, Rev. 32, ESF Power Availability

4. Drawing OOOOOEOAAAA, Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. 1&2 133 P1 a o e SS5 ECL: SITE AREA EMERGENCY Initiating Condition:

Inability to shutdown the reactor causing a challenge to core cooling or RCS heat removal.Operating Mode Applicability: 1, 2 Emergency Action Level: (1) a. An automatic or manual trip did not shutdown the reactor.AND b. ALL manual actions to shutdown the reactor have been unsuccessful. AND c. EITHER of the following conditions exists: " Core Cooling -Red entry conditions met OR* Heat Sink- Red entry conditions met Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a SITE AREA EMERGENCY. In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting firom an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a SITE AREA EMERGENCY in response to prolonged failure to shutdown the reactor.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC RGIRG 1 or FG1.SS5: EAL-1 Selection Basis: Core Cooling -Red entry conditions met (CETs > 12000 F) is the site specific indication of the inability to adequately remove heat from the core. Heat Sink -Red entry conditions met (NR level in All SG <14% [34%] AND total AFW flow to SG < 576 GPM) is the site specific indication of the inability to remove heat from the RCS.134 1 P a g e

REFERENCES:

1. Procedure OPOP05-EO-F002, Rev. 2. Core Cooling Critical Safety Function Status Tree 2. Procedure OPOP05-EO-F003, Rev. 6, Heat Sink Critical Safety Function Status Tree 1351P age SS8 ECL: SITE AREA EMERGENCY Initiating Condition:

Loss of ALL Vital DC power for 15 minutes or longer.Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Level: Note: The Emergency Director should declare the SITE AREA EMERGENCY promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Indicated voltage is less than 105.5 VDC on ALL Class IE 125 VDC battery buses for 15 minutes or longer.Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In miodes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via ICs RG1, FGI or SG8.SS8: EAL-1 Selection Basis: Minimum voltage for Class IE 125 VDC battery buses was determined in calculation 13-DJ-006 Rev.3 and determined to be 105.5 volts. At 105.5 volts or less, OPOP05-EO-ECOO, Loss of All AC Power directs the operators to open the battery output breakers.

REFERENCES:

1. OPOP05-EO-ECOO, Rev. 23, Loss of All AC Power 1361 IP a ( e SG1 ECL: GENERAL EMERGENCY Initiating Condition:

Prolonged loss of ALL offsite and ALL onsite AC power to emergency buses.Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Level: Note: The Emergency Director should declare the GENERAL EMERGENCY promptly upon determining that 4 hours has been exceeded, or will likely be exceeded.(1) a. Loss of ALL offsite and ALL onsite AC power to ALL three 4160V AC ESF Buses.AND b. EITHER of the following:

• Restoration of at least one 4160VAC ESF bus in less than 4 hours is not likely.* Core Cooling -Red entry condition met Basis: This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers.

In addition, fission product barrier monitoring capabilities may be degraded under these conditions. The EAL should require declaration of a GENERAL EMERGENCY prior to meeting the thresholds for IC FG 1. This will allow additional time for implementation of offsite protective actions.Escalation of the emergency classification from SITE AREA EMERGENCY will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of four (4) hours. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is a higher likelihood of challenges to multiple fission product barriers.The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.The EAL will also require a GENERAL EMERGENCY declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.137 1 P a .e SGI: EAL-1 Selection Basis: The prolonged loss of all onsite and all offsite AC power coupled with Core Cooling -Red entry conditions (CETs > 12000 F) are sufficient indications of the inability to remove heat from the core.Station Blackout does not include the loss of available AC power to buses fed by station batteries through inverters, or by Alternate AC (AAC) sources as defined in NUMARC 87-00. The STPEGS Station Blackout position credits any one of the three Standby Diesel Generators as the AAC source. The required coping duration category determined for STPEGS Station Blackout is a minimum of four hours, based on the guidance of NUMARC 87-00, Section 3. STPEGS meets this requirement and forms the basis for the four hour time period.

REFERENCES:

1. 0POP04-AE-0001, Rev. 44, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus 2. OPOP04-AE-0004, Rev. 15, Loss of Power to One of More 4.16 KV ESF Buses 3. OPSP03-EA-0002, Rev. 32, ESF Power Availability

4. Drawing OOOOOEOAAAA, Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. 1&2 5. OPOP05-EO-F002., Rev. 2, Core Cooling Critical Safety Function Status Tree 6. OPOP05-EO-ECOO, Rev. 23, Loss of All AC Power 7. STPEGS UFSAR Section 8.3.4, Station Blackout 138 [P a a e SG8 ECL: GENERAL EMERGENCY Initiating Condition:

Loss of ALL AC and Vital DC power sources for 15 minutes or longer.Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels: Note: The Emergency Director should declare the GENERAL EMERGENCY promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. Loss of ALL offsite and ALL onsite AC power to ALL three 4160V AC ESF buses for 15 minutes or longer.AND b. Indicated voltage is less than 105.5 VDC on ALL Class 1E 125 VDC battery buses for 15 minutes or longer.Basis: This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fiuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.SG8: EAL-1 Selection Basis: This IC and EAL were included to address the operating experience for the March, 2011 accident at Fukushima Daiichi. Minimum voltage for Class 1E 125 VDC battery buses was determined in calculation 13-DJ-006 Rev.3 and determined to be 105.5 volts. At 105.5 volts or less, 0POP05-EO-EC0O, Loss of All AC Power directs the operators to open the battery output breakers.

REFERENCES:

1. OPOP04-AE-0001, Rev. 44, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus 2. OPOP04-AE-0004, Rev. 15, Loss of Power to One of More 4.16 KV ESF Buses 3. OPSP03-EA-0002, Rev. 32, ESF Power Availability

4. OPOP05-EO-EC0O, Rev. 23, Loss of All AC Power 5. Drawing OOOO0EOAAAA, Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. 1&2 139 [ P a -e APPENDIX A -ACRONYMS AND ABBREVIATIONS AC .................................................................................................................................

Alternating Current AOP ........................................................................................................... Abnormal Operating Procedure ATW S .............................................................................................. Anticipated Transient W ithout Scram CDE ................................................................................................................ Com m itted Dose Equivalent CFR ................................................................................................................ Code of Federal Regulations CTM T/CNM T ......................................................................................................................... Containm ent CSF ........................................................................................................................ Critical Safety Function CSFST ................................................................................................. Critical Safety Function Status Tree DBA ......................................................................................................................... Design Basis Accident DC ......................................................................................................................................... Direct Current EAL ...................................................................................................................... Emergency Action Level ECCS ...................................................................................................... Em ergency Core Cooling System ECL ........................................................................................................... Emergency Classification Level EOF ............................................................................................................ Emergency Operations Facility EOP .......................................................................................................... Emergency Operating Procedure EPA ........................................................................................................ Environmental Protection Agency EPG ......................................................................................................... Emergency Procedure Guideline ERG .......................................................................................................... Emergency Response Guideline FEM A ........................................................................................ Federal Emergency M anagem ent Agency FSAR ............................................................................................................. Final Safety Analysis Report GE .................................................................................................................... GENERA L EM ERGENCY IC ................................................................................................................................. Initiating Condition ID ....................................................................................................................................... Inside D iam eter ISFSI ...................................................................................... Independent Spent Fuel Storage Installation Keff .............................................................................................. Effective Neutron M ultiplication Factor LCO ......................................................................................................... Lim iting Condition of Operation LOCA .................................................................................................................. Loss of Coolant Accident M SIV .............................................................................................................. M ain Steam Isolation Valve M SL ................................................................................................................................. M ain Steam Line mR, toRem , m rem , mREM ....................................................................... m illi-Roentgen Equivalent M an M W .............................................................................................................................................. M egawatt NEI ........................................................................................................................ N uclear Energy Institute NPP ............................................................................................................................ N uclear Power Plant NRC ......................................................................................................... Nuclear Regulatory Com m ission NSSS ........................................................................................................... Nuclear Steam Supply System NORA D ............................................................................ North American Aerospace Defense Com mand (NO )UE ................................................................................................... (Notification Of) Unusual Event NUM ARC ............................................................................ N uclear M anagement and Resources Council OBE ................................................................................................................ Operating Basis Earthquake OCA ....................................................................................................................... Owner Controlled Area ODCM .................................................................................................... Offsite Dose Calculation M anual ORO ........................................................................................................... Off-site Response Organization PA ........................................................................................................................................ Protected Area PAG ................................................................................................................ Protective Action Guideline PRA/PSA,.......................... .Probabilistic Risk Assessment / Probabilistic Safety Assessment 1401Page PW R ................................................................................................................... Pressurized W ater Reactor PSIG ............................................................................................................ Pounds per Square Inch Gauge R .................................................................................................................................................... Roentgen RCS ....................................................................................................................... Reactor Coolant System Rein, rein, REM ................................................................................................. Roentgen Equivalent M an RPS ................................................................................................................... Reactor Protection System RPV ........................................................................................................................ Reactor Pressure Vessel RVW L ............................................................................................................. Reactor Vessel W ater Level SAR ......................................................................................................................... Safety Analysis Report SCBA ................................................................................................. Self-Contained Breathing Apparatus SG ..................................................................................................................................... Steam Generator SI ........................................................................................................................................ Safety Injection SPDS ....................................................................................................... Safety Parameter Display System TEDE ........................................................................................................ Total Effective Dose Equivalent TOAF ............................................................................................................................. Top of Active Fuel TSC ..................................................................................................................... Technical Support Center W OG ............................................................................................................. W estinghouse Owners Group 1411 Pao e APPENDIX B -DEFINITIONS The following definitions are taken from Title 10, Code of Federal Regulations, and related regulatory guidance documents. ALERT: Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels.GENERAL EMERGENCY: Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.UNUSUAL EVENT UE: Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.SITE AREA EMERGENCY: Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; I) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.The following are key terms necessary for overall understanding the emergency classification scheme.EMERGENCY ACTION LEVEL (EAL): A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given EMERGENCY CLASSIFICATION LEVEL.EMERGENCY CLASSIFICATION LEVEL (ECL): One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1)potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The EMERGENCY CLASSIFICATION LEVELS, in ascending order of severity, are:* UNUSUAL EVENT UE* ALERT" SITE AREA EMERGENCY (SAE)" GENERAL EMERGENCY (GE)FISSION PRODUCT BARRIER THRESHOLD: A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.142 P age INITIATING CONDITION (IC): An event or condition that aligns with the definition of one of the four EMERGENCY CLASSIFICATION LEVELS by virtue of tile potential or actual effects or consequences. Selected terms used in INITIATING CONDITION and EMERGENCY ACTION LEVEL EMERGENCY ACTION LEVEL statements are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below.CONFINEMENT BOUNDARY: The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage.CONTAINMENT CLOSURE: Those actions necessary to place the RCB in the closed containment condition that provides at least one integral barrier to the release of radioactive material. Sufficient separation of the containment atmosphere from tile outside environment is to be provided such that a barrier to the escape of radioactive material is reasonably expected to remain in place following a core melt accident.CREDIBLE SECURITY THREAT: Information received from a source determined to be reliable (e.g., law enforcement, government agency, etc.) or has been verified to be true or considered credible when: (1) Physical evidence supporting the threat exists, (2) Information independent from the actual threat message exists that supports the threat, or (3) A specific known group or organization claims responsibility for the threat.EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressuLrization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present.FACILITY: The area and buildings within the PROTECTED AREA and the switchyard. FAULTED: The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.HATCH MONITOR: Temporary monitor installed when Containment High Range Radiation Monitors RT-8050 and RT-8051 are out of service.HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station.143 11P a -e HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.NORMAL LEVELS: As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value.OWNER CONTROLLED AREA: The area surrounding the PROTECTED AREA where STP Nuclear Operating Company (STPNOC) reserves the right to restrict access, search personnel, and vehicles.PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.PROTECTIVE ACTION GUIDES (PAG): Environmental Protection Agency (EPA) guides for protective actions to safeguard against radiation exposure from nuclear incidents. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan.REFUELING PATHWAY: Includes all the cavities, tubes, canals and pools through which inradiated fuel may be moved, but not including the reactor vessel.RUPTURE(D): The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition., including the ECCS. These are typically systems classified as safety-related SECURITY CONDITION: Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.144 1P a u.e SECURITY EVENT: Any incident representing an attempted, threatened, of actual breach of the security system of reduction of the operational effectiveness of that system. A security event can result in either a SECURITY CONDITION or HOSTILE ACTION.SITE BOUNDARY: The edge of the plant property whose access may be controlled by STPEGS. This boundary is congruent with the Exclusion Area Boundary for the purpose of offsite dose assessment. UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally.UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.THYROID CDE: The dose equivalent to the thyroid from an intake of radioactive material by an individual during the 50-year period following the intake.VALID: An indication, report or condition is considered to be VALID when it is verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. This may be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. The verification methods should be completed in a manner the supports timely emergency declaration. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. 145 1 P a g e Attachment 3 STPEGS Emergency Action Level Technical Bases Document -redline markup NEI 99-40 6]STPEGS Emergency Action Level Technical Bases Document Rev. 0 hruI , QW.0 Raw' 61 ,mimnator~ !!-U+/- Key. b Implementation I~-aa vn entn November 201 &T-TOMVJ1 Fet--7y 5 NOTE: Changes to this document require a review under 1OCFR50.54 (q) as directed by OPGP05-ZV-0010, Emergency Plan Change. m a~o' HEo. was PreParEc 85' 1v in'oeicaf rcrg," ins o c.P'1! tL. rFgeonc' ACtO, LEeil ttLAL) Is aP- Poree.NEI Chniorpcr-sa: David Younig Preparatfico TeRam La..m,' Bak..r .e.lon Nuclear./Co.pot..e C-raig B .anne. r 2 -PSG and Hope Cr-eek Nuclear Generating Stations/USA John Egdorf ...... an Power Station Jack Lewis Entergy NuElear-'/CoEperate C. Kelly W.alk.. r Oprations Suppon. N6e.Rev~ew- eam ChisBone Southom Nucleor/Corporole Bill Chal-sse Encrooen Sorv~icee, hie.KenHt Crooker-- PrFogeFss EnRoo'B9Frunewie-Nuclolar Plant Don Crow.l D-ue Energy/C-orporate Rogor Freontian Cons~tellto Enr- Nuolar Group/CorPorate W.A,7alt ... Lo A Nuclear/Co rporato Ken Meade FENOC/Corporate Don Mothena NctkrAFa Energy/Coporate-David StEbaugh! EP Consulting, LLC-a-ee --a--i- Diabloeano Powor Plant/STARS NOTICE-".'-", !io, a of its itpioyeec, mntlrors, org.ani.ation, contracts, or consult.ant. s ntai any wam.an... ..XPcS.eE. Or iIpiici, or assume any legal responsibili y for the accuracy or co..pletnc.s. of... or assume any liab ity. fo damages resulting ron any use ofý,are;inauto caraRtus. A;@thods. or RCroces disclosed4 in thi reoq o-rA tFhat such ma notA infinc crivamely o'emd riohts. Federal rgltosrequire that a ncldear po)wer planlt operator develop a se3herne fcrF the- classificatil on teioergency events And eAnlditions. This seherne is an funldamental componient ofan emergency plan inthat itprovides the eie thresqholds thlat vill alias% Site personnel to rapidlyf im~pl~emeiita angeeOf PH ,ie4i~ed eflwffgelizff pOnSP)e mleasur-es. Ani emerfgency classification scheme also theilitates tinmcl'; decisiei ntakine by an Offite Res.pons..e Org..an ion (ORO). cnc.ernin.g the im..plemen. ."tat oprcautionary or protestive acin for the public.The purpese of Nuclear Energy Institute (NEI) 99 01 is to provYide guEidancee to nuclear p.owNer plantl opertors far the 'developmenit of asite speeifi emfffergenc classification schemfe. The methaodaky described in this dOeumlrt iS Eeon itInt vAith Federal regulationls, and relatedi U-S Nuclear Regulatory Commission (NRCIF~~ requireents and gui dance. In particular, this me~tho~dology has been! endorsed by the NRC as an lecepieble-aPP!Oih to Meetinig thFe requirementsef 10 CFR § 2 740. related sections of 10 CFR 5App , plann ing stanldard evaluaFt ion elements ofNUREC 06A WI [PEA REP 1, Rev. I, iwj~ CWof~pcsciaa Ervalhwst/or oqflafo/dogio! .411eitgoem: Resp'onse Pious Hif Po'dP:ePa;d;oess i;; S* P. :', ro" oer P.. ..s, No.em.be W8. o NPt 99 01 oonbinsm a set of g@-eneri InitifltiogL C-ondiOIlL (0), Effergency Action Lev ela (FAIs) and fissionl product barrier status thresholds. it AsO includes supporting technical basis information, developer notes and reEomm-e.nded classification instr-cti'Os for usFers. User.s ,should i1plement ... , AE and t.iheshRd,. that ore a: as P b LI t"ithe trial presented... in......... thicdocum.ent.. 1 it; Vith allowa.Hn.e for chan..s necessan; to address site specific consideratossuc h as plant design, lti.. terminolog.. etc.Propery implemtented, the guid..t..ee i NEI 99 .' mill t i.d.a site specific energency classification she..me with cleary defined and readily bSer.'able PA~ an Fd thresholds. OtherF benelf~its inlude the developtoent of a Sound basis docuenllit, thle adoptOian Of ilidulstfr stanldard inStruc.tions for1 .emerg.-ency.c 'lassificat in (e.g., tran.sient events. classification of.multiple ev ents, upgrf.ading., g rig.... e .incorpation of features to improve human performance. An emergency c.,..lassiOicatio using this schem .sill be appropriate to the risk pod to plant w.okers and the pulblic. and should he the same aS that mo-de- .' ISOh 919 01i user plant in respooe4;' toa sImia event.The indiv'iduals responsible for g an .emerg.eny classic , atio n ScheM are st.roly encoaged to review all applicable NRC requirement.an. d guidance prior to beginning their effots. Questions concerning this document may he directed t0 the NE! E:ms rgency Preparedness staff- NEl EAL task force mfemfbers or subminied to the Emergency PreparetesFeqetyAkedl-Qulestions proeess.Finally , uniqule State and local reqireet asscited With ani emlergency classifiation; Schem are; no Rt reýflected in this guidance. ineol~orationi Of .t"he requiremen.ts may be perform-ed , on a eas-b cby .asL basis in conjunction 'vith popriate ORO .A.ny such changes W.ill require are. ie',v under the applicable sections of 10 C. R I 0. [l Iii6 0 ItNj .:0 LIfF'1k I , [iL !Nr N IN I EIN I IUk. NA 1,L[. YV] TABLE OF CONTENTS EXECUTIVE

SUMMARY

.......................................................................................................

1 I REGULATORY BACKGROUND ................................................................... 1 DEVELOPMENT OF EMERGENCY ACTION LEVELS ........................................................ 1 1.1 O PE R A T IN G RE A C T O R S ................................................................................................................ 1 1.1 REG U LA TO R Y B A C K G RO UN D .................................................................................................... 1 1 D2 PERMANENTLY DEFUELED STATION i 1.3 I3N) S.PENT FU .EL-T STOPA 6 .GE I NsTTA11 AT I ON (Is FSI) 1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ..................................... 2 1.i N R C O R D ,ER E A 12 051 ................................................................................................................... 2 1.3 N RC O RD ER EA 051 ............................... ...................................... 3 1.5 APPLICABILITY TO ADVANCED ADD SMALL MODULAR EACTRS .............................. 1 2 KEY TERMINOLOGY USED IN NEI 99-01............... .. ................. 4 2 KEY TERM INOLOGY ......................................................................................................... 4 2.1 EMERGENCY CLASSIFICATION LEVEL (ECL) ....................................................................... 5 2.2 IN ITIATIN G CON D ITIO N (IC) .................................................................................................. 6 2.3 EM ERGENCY ACTION LEVEL (EAL) ........................................................................................ 6 2.4 FISSION PRODUCT BARRIER THRESHOLD ........................................................................... 7 3 DESIGN OF THE NEI 99-01 EMERGENCY CLASSIFICATION SCHEME .............. 7 3 DESIGN OF THE STPEGS EMERGENCY CLASSIFICATION SCHEME ............................ 8 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS) ................................ 8 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS ..................... 11 3.3 NSSS DESIGN DIFFERENCES ........................................................................... i13 3.3 STPEGS DESIGN CON SIDERA TIONS ..................................................................................... 11 3.4 ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION ........................... 12 3.5 IC AND EAL MODE APPLICABILITY .......................................... 13 4 SITE SPECIFIC SCHEME DEVELOPMENT GUIDANCE ......................... 4 STPEGS SCHEME DEVELOPMENT ............................................................................ 15'1.1 GENEIRAL4 IM4PLEEN4sTATION GUIDANCE I1Q'!.!~~ ~~~~~ .EE .L ! P E E 'T 'T O N G I A C ........................................................................... -- -4.1 GENERAL DEVELOPM ENT PROCESS ................................................................................... 15 4.2 CRITICA L CHA RA CTERISTICS ................................................................................................ 16 4.3 INSTRUM ENTATION USED FOR EALS ................................................................................. 17 4.4 PRESENTATION OF SCHEME INFORM.ATIO.. ..TO USER ....................................................... 20'1.5 INTEGRATION OF ICSiEALS WITH PLANT PROCEDURES...................................... 21 41.6 BASIS DOCUMENT .............................................. .............

4.7 DEVELOPER

AND USER. FEED..AC.

• • --'*....

....

4.4 REFERENCES

TO STPEGS AOPS AND EOPS-. ....................................................................... 19 5 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ........................................ 19 5.1 GENERAL CONSIDERATIONS ........................... ...................................................... 19 5.2 CLASSIFICATION M ETHODOLOGY ..................................................................................... 20

5.3 CLASSIFICATION

OF MULTIPLE EVENTS AND CONDITIONS ....................................... 20 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION ............................. 21 5.5 CLASSIFICATION OF IMMINENT CONDITIONS ". 1...........................21

5.6 EMERGENCY

CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING ............ 22 5.7 CLASSIFICATION OF SHORT-LIVED EVENTS ............................ ....................................... 22 5.8 CLASSIFICATION OF TRANSIENT CONDITIONS ................................. ........... 22 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION ....... 23 5.10 RETRACTION OF AN EMERGENCY DECLARATION ...................................................... 23 6 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/EALS .......................... 24 7 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS ................. 5455 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS ............. 899 9 FISSION PRODUCT BARRIER ICS/EALS ................................................... ............... 9293 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS .. 128I-,O 11 SYSTEM MALFUNCTION ICS/EALS .................................................................... 1601-62 APPENDIX A -ACRONYMS AND ABBREVIATIONS ................................................ 200202 APPENDIX B -DEFINITIONS ................................................................................... 202204 APPENDIX C -PERMANENTLY DEFUELED STATION ICs/EALs. 193 THIS PAGE IS LEFT INTENTIONALLY BLANK DEVELOMENT OF EMERGENCY ACTION LEVELS FOR NON-PASSIVE REACTORS 1 REGULATORY BACKGROUND 1 DEVELOPMENT OF EMERGENCY ACTION LEVELS 1. 1 OPERATING PEACTORS 1.1 REGULATORY BACKGROUND Title 10, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC)regulations that apply to nuclear power facilities. Several of these regulations govern various aspects of an emergency classification scheme. A review of the relevant sections listed below will aid the reader in understanding the key terminology provided in Section 3.0 of this document.* 10 CFR § 50.47(a)(1)(i)

• 10 CFR §'50.47(b)(4)
• 10 CFR § 50.54(q)* 10 CFR § 50.72(a)* 10 CFR § 50, Appendix E, IV.B, Assessment Actions* 10 CFR § 50, Appendix E, IV.C, Activation of Emergency Organization The above regulations are supplemented by various regulatory guidance documents.

Three documents of particular relevance to NEI 99-01 are:* NUREG-0654/FEMA-REP-I, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants]* NUREG-1022, Event Reporting Guidelines 10 CFR § 50.72 and § 50.73--Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactor" The above list is not all inclusive and it is strongl r.ecommended that scheme developer-s ... .nsult with licensing/regulatory complianec per-sonnel to idcntiby and uinderstand all applic-able requtirementsan guidance. may als..... bhe ddi.racte.ed to the NEI Emergencv Prepar'edness staff.1.2 PERMANENTLY DEFUELED STA TION NEI 99 01 proevides guidance for an emergIency classification seheme applicable to a permanently defuieled station. This is a station that generated spent fuel under a 10 CFR § 50 license, has permaniently cease operations and- w.ill stare the spent fuel onsite for an extended period of time. The emer-gency classific-ation levels I Wpage applicable k) this ty.pe of StatiOn _ae cAontn Wi the., requirements of 10 DFR § 50 and the guidance in X1ITI Cf- E %C-CA IE'I, A A D UD I in order te relax the emergency plan requiremants applicable to an operating statin, the owner of a permfanently defueled station must demn!Ostrate that no crdible event can result in a significant radiological r-elease beyond the site boundary. it is expeted that this erSTPificSai will conwfirm that the source tedrea and motive fce available in the perimanenly defeled ofndition aFre insufficient to warrant classifications of a Site Afiea Emegeney or General Emergency. Therefore, the generic initiating Conditions (!Gs) and emergency Action Lasvels (EALs)applicable to a peirmanently defueled station miay r-esult in either a Notification of Unusual Event (NOUE) or anl Aler1 classif6iation. The geineric Ws and EALs are pfesented in Appendix C, PeIFaSIenly DIfueled Station Is/EALs.1.3 Th4DEPENDENT SPENT FUELn STORAGE PNSTALLJATION (ISFSI)4-41.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFST)South Texas Proiect Electrical Generating Station (STP or STPEGS) is locating anl ISFSI approximatel450 feet west ofthe Unit 2 Reactor Building. The STP ISFSI will be within the site Protected Area and is scheduled to be operational in 2016.Selected guidance in NEI 99-01 is applicable to the STPEGS licensees electing, to usel10 CER 50 emergency plan to fulfill the requirements of 10 CFR 72.32 for a stand-alone ISFSI. The eSergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CFR § 50 and the guidance in NUREG 0654/FEMA-REP-l. The initiating conditions germane to a 10 CFR § 72.32 emergency plan (as described in NUREG-1 567) are Subsumed within the classification schene for a 10 CFR § 50.47 emergency plan.The ge+i@Fe- STPEGSICs and EALs for an ISFSl are presented in Section 8, ISFSI ICs/EALs. IC E-HU I covers the spectrum of credible natural and man-made events included within the scope ofoat -the STPEGS ISFSd I design.. This IC is not applicable to installations or fclitathat ma:d prcess andr repaclage spent fuel (e.g., a moaitmred Retrievable Stoaage Faeility orf an pSFSi at a spent ftel preeessine i addition, appropriate aspects of IC HUI and IC HAl sheuld also be ineluded to address a HOSTILE ACTION directed against a-the STPEGS ISFSI.The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-l 140, A RegilatoyjAnal'sis on Emnergenci.' Pr-epar-edness for- Futel C'ycle and Othe1 r Radioactive Al.atergial Licensees. NeREG-l 140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials Would not exceed I rem Effective Dose Equivalent. Regarding the above information, the expectations for an offsite response to an Al-ert-ALERT classified under a 10 CFR § 72.32 emergency plan are generally consistent with those for an-a Notification of Uusuqal Even.1t UNUSUAL EVENT in a 10 CFR § 50.47 emergency plan (e.g., to provide assistance if requested). Also, the lieensee's STPEGS Emergency Response Organization (ERO) required fora 10 CFR § 72.32 emergency plan is different than that prescribed for a 10 CFR § 50.47 emergency plan (e.g., no emergency technical support function). 2 P a g e 1.4 NRT ORDER EA 12 051-1.3 NRC ORDER EA-12-051 The Fukushimna Daiichi accident of March 11,2012, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity, and ultimately led to core damage in three reactors. While the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to preclude fuel damage from the loss of cooling.Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii). Among them was to provide each spent fuel pool with reliable level instrumentation to significantly enhance the ability of key decision-makers to allocate resources effectively following a beyond design basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to Modi, Licenses with Regard to Reliable Spent Fuel Pool Instrunentation, on March 12, 2012, to all US nuclear plants with an operating license, construction permit, or combined construction and operating license.NRC Order EA-1 2-051 states, in part, "All licensees ... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel: (1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." To this end, all licensees must provide: " A primary and back-up level instrument that will monitor water level from the nornmal level to the top of the used fuel rack in the pool;* A display in an area accessible following a severe event; and* Independent electrical power to each instrument channel and provide an alternate remote power connection capability. NEI 12-02. Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", provides guidance for complying with NRC Order EA-I12-051. NEI 99 01 .Revision 6, This document includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-051. These EALs are included within existing IC A-ARA2. and new ICsAS RS2 and-AG- RG2. It is reco:mmr. ended that tThese EALs will be implemented when the enhanced spent fuel pool level instrumentation is available for use.The regulatory process that licensees follow to make chianges to their emiergency plan, including non schem~e changes to EALs. is 10 CER 50.54(q). in acco-erd-Ance vvith this regulation. licensees are responsible for evaluating" reposed change and detefmining-whether: ori not it r-esults in a reductioni in the effccetiveniess of the plani. As a....... -..... ..a.e 3[1P agRe

1.5 APPLICABILITY

TO ADVANCE AND SMALL MAODULAR REACTO DERCIQN The gaidance in this docum~ent primarily add ese emi~ercial nuelear pow;Aer reacters in the United States, opei-atiflg Or permfaaently defuceled. as of 2012 (so called lz an 2n geni@Fatiain plant designis): hev.ever. it may b adapted to advan..ed non passive designs (often r.efcrred o as 3fg ....rati... I .l desig.s) as well. Developers o ani emfer-gency classification scheme for- an advanced non passive reaetor- plant may need to proapose deviations froem the generici guidance to account for- the differences inq design par-ameters and criteria, and operating characteristics and capabilities, between 2nJand-3 rd generation plants.Thoce are significant design and operating difaferenes between lage commerial nuclear power plants (Mf any genefration) and Small Modular- Reactor-s (Wk~s) (e.g., differences in source term). For this r-easoni,-4this documient is not aeolieable to SMRs.ucy EIvnnpn nnnIULfl flE iagnE Eim mE~ CO Al-.p * * *u*um* B nww -.w *

• U
• u* wa 2 KEY TERMINOLOGY USED There are several key terms that appear throughout the NEI 99 01 EAL methodology.

These tenns are introduced in this section to support understanding of subsequent material. As an aid to the reader, the following table is provided as an overview to illustrate the relationship of the terms to each other............. Level .EMERGENCY CLASSIFICATION LEVEL AIei4ALERT SAE GE-veetUNUSUAL EVENT+ 40 Initiating Condition Initiating Condition Initiating Condition Initiating Condition+ .4 0 Emergency Action Emergency Action Emergency Action Emergency Action Level (1) Level (1) Level (1) Level (1)" Operating Mode -Operating Mode

• Operating Mode
• Operating Mode Applicability Applicability Applicability Applicability" Notes
• Notes
• Notes
• Notes" Basis
• Basis
• Basis
• Basis (1) -When making an emergency classification, the Emergency Director must consider all infornation having a bearing on the proper assessment of an Initiating Condition.

This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the infonning Basis information. In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL.4 1P a ge --EMERGENCY CLASSIFICATION LEVEL (ECL)4-22.1.One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification leveLEMERGENCY CLASSIFICATION LEVELS, in ascending order of severity, are:* Notifleatien of Uaustual Eve4tUNUSUAL EVENT (-NOQU&LUE) SAei4-ALERT

• Site Area EmergeneySITE AREA EMERGENCY (SAE)* General EmergencGENERAL EMERGENCY (GE)2.1.1 2. 1.lNotificaticn ofUinu, ,il Event UNUSUAL EVENT (NO4EUE)Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.Purpose: The purpose of this classification is to assure that the first step in future response has been carried out, to bring the operations staff to a state of readiness, and to provide systematic handling of unusual event information and decision-making. 2.1.2 2.1.2Aeirt ALERT Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels.Purpose: The purpose of this classification is to assure that emergency personnel are readily available to respond if the situation becomes more serious or to perform confirmatory radiation monitoring if required, and provide offsite authorities current information on plant status and parameters. 2.1.3 2.1.3Site Area Emergen...SITE AREA EMERGENCY- (SAE)Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.Purpose: The purpose of the Site Area En-ergenevSITE AREA deelarationEMERGENCY declaration is to assure that emergency response centers are staffed, to assure that monitoring teams are dispatched, to assure that personnel required for evacuation of near-site areas are at duty stations if the situation becomes more serious, to--.. .. -.-.--. .-.---.. .. ... .. ... .... ..... ....................I-P a a e provide consultation with offsite authorities, and to provide updates to the public through government authorities. 2.1.4 2.4.,Ceneral Emtergeney GENERAL EMERGENCY -(GE)Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.Purpose: The purpose of the General Emergency GENERAL EMERGENCY declaration is to initiate predetermined protective actions for the public, to provide continuous assessment of information from the licensee and offsite organizational measurements, to initiate additional measures as indicated by actual or potential releases, to provide consultation with offsite authorities, and to provide updates for the public through government authorities. --INITIATING CONDITION (IC)4-,32.2 An event or condition that aligns with the definition of one of the four emergency classification leveLs EMERGENCY CLASSIFICATION LEVELS by virtue of the potential or actual effects or consequences. Discussion: An IC describes an event or condition, the severity or consequences of which meets the definition of an emergency classification level. An IC can be expressed as a continuous, measurable parameter (e.g., RCS leakage), an event (e.g., an earthquake) or the status of one or more fission product barriers (e.g., loss of the RCS barrier).Appendix I of NUREG-0654 does not contain example Emergency Action Levels (EALs) for each ECL, but rather Initiating Conditions (i.e., plant conditions that indicate that a radiological emergency, or events that could lead to a radiological emergency, has occurred). NUREG-0654 states that the Initiating Conditions form the basis for establishment by a licensee of the specific plant instrumentation readings (as applicable) which, if exceeded, would initiate the emergency classification. Thus, it is the specific instrument readings that would be the EALs.Considerations for the assignment of a particular Initiating Conditicn INITIATING CONDITION to an emer-gency classifieation le-el EMERGENCY CLASSIFICATION LEVEL are discussed in Section 3.2.2.1 EMERGENCY ACTION LEVEL (EAL)A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level.Discussion: EAL statements may utilize a variety of criteria including instrument readings and status indications; observable events; results of calculations and analyses; entry into particular procedures; and the occurrence of natural phenomena. 61Pa e 2.2.2 2-.4FISSION PRODUCT BARRIER THRESHOLD A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.Discussion: Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers., any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are:* Fuel Clad* Reactor Coolant System (RCS)* Containment Upon determination that one or more fission product barrier thresholds have been exceeded., the combination of barrier loss and/or potential loss is compared to the fission product barrier IC/EAL criteria to determine the appropriate ECL.In some accident sequences, the ICs and EALs presented in the Abnormal Radiation Levels/ Radiological Effluent (RA) Recognition Category will be exceeded at the same time, or shortly after, the loss of one or more fission product barriers. This redundancy is intentional as the former ICs address radioactivity releases that result in certain offsite doses from whatever cause, including events that might not be fully encompassed by fission product barriers (e.g., spent fuel pool accidents, design containment leakage following a LOCA, etc.).71Page 3 DESIGN OFTHE NEI 99-01 EMERGENCY CLASSIFICATION 3 DESIGN OF THE STPEGS EMERGENCY CLASSIFICATION SCHEME-ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS)243.1 An effective emergency classification scheme must incorporate a realistic and accurate assessment of risk, both to plant workers and the public. There are obvious health and safety risks in underestimating the potential or actual threat from an event or condition; however, there are also risks in overestimating the threat as well (e.g., harm that may occur during an evacuation). The NEI 99 01 emergency classification scheme attempts to strike an appropriate balance between reasonably anticipated event or condition consequences, potential accident trajectories, and risk avoidance or minimization. There are a range of "non-emergency events" reported to the US Nuclear Regulatory Commission (NRC) staff in accordance with the requirements of 10 CFR § 50.72. Guidance concerning these reporting requirements, and example events, are provided in NUREG-1022. Certain events reportable Under the provisions of 10 CFR §50.72 may also require the declaration of an emergency. In order to align each Initiating Conditions (IC) with the appropriate ECL, it was necessary to determine the attributes of each ECL. The goal of this process is to answer the question, "What events or conditions should be placed Linder each ECL?" The following sources provided information and context for the development of ECL attributes." Assessments of the effects and consequences of different types of events and conditions

• STPEGSTypic*

abnormal and emergency operating procedure setpoints and transition criteria* STPEGSTypii4e Technical Specification limits and controls* al Efflbuiett T Specifications (RETS)!STPEGS Offsite Dose Calculation Manual (ODCM)radiological release limits* Review of selected STPEGS Updated Final Safety Analysis Report (UFSAR) accident analyses* Environmental Protection Agency (EPA) Protective Action Guidelines (PAGs)* NUREG 0654, Appendix I, Emergency Action Level Guidelines for Nuclear Power Plants I lndustry Operating Experience

• Input from ind'!Stf+'

subject matter experts and NRC staff'membersat STPEGS The following ECL attributes were created by the Revision 6 Preparation Team; to aid in the development of ICs and Emergency Action Levels (EALs). The team decided to include the attributes in this revision since they The attributes may be useful in briefing and training settings (e.g., helping an Emergency Director understand why a particular condition is classified as an Aei4ALERT. It should be stressed that developers not at.empt to redefin.these a.ributes ,, apply them in any fashi.n that would .hange the generic guidance contained in this document.8 1P a Re The attributes of each ECL are presented below.-2443. I. 1 3.1.1 NOtifi .atiOn ofl' U.u..al Eveint UNUSUAL EVENT (NO4JUIE)AnA Notification of Ukuual Event UNUSUAL EVENT, as defined in section 2.1.1, includes but is not limited to an event or condition that involves: (A) A precursor to a more significant event or condition.(B) A minor loss of control of radioactive materials or the ability to control radiation levels within tile plant.(C) A consequence otherwise significant enough to warrant notification to local, State and Federal authorities. 2.1.23. 1.2 3..2 Alet ALERT An A-ei4ALERT, as defined in section 2.1.2, includes but is not limited to an event or condition that involves: (A) A loss or potential loss of either the fuel clad or Reactor Coolant System (RCS) fission product barrier.(B) An event or condition that significantly reduces the margin to a loss or potential loss of the fuel clad or RCS fission product barrier.(C) A significant loss of control of radioactive materials resulting in an inability to control radiation levels within the plant, or a release of radioactive materials to the environment that could result in doses greater than 1% of an EPA PAG at or beyond the site boundary.(D) A HOSTILE ACTION occurring within the OWNER CONTROLLED AREA. ine,"dilig !hce difRccted At An lindependenit SPOnt 14uel Stor-ag@ installation (ISESI).-24_3.1 .3.1.3 Site Ar'ea EmergeneySITE AREA EMERGENCY- (SAE)A Site Area EmnlgeRneySlTE AREA EMERGENCY, as defined in section 2.1.3., includes but is not limited to an event or condition that involves: (A) A loss or potential loss of any two fission product barriers -fuel clad, RCS and/or containment.(B) A precursor event or condition that may lead to the loss or potential loss of multiple fission product barriers within a relatively short period of time. Precursor events and conditions of this type include those that challenge the monitoring and/or control of multiple safety systems.(C) A release of radioactive materials to the environment that could result in doses greater than 10% of an EPA PAG at or beyond the site boundary.(D) A HOSTILE ACTION occurring within the plant PROTECTED AREA.9 1P a ge 24433.1.4 -3.1.4 General Emer'gency GENERAL EMERGENCY- (GE)A General Emer'gency.GENERAL EMERGENCY. -as defined in section 2.1.4., includes but is not limited to an event or condition that involves: (A) Loss of any two fission product barriers AND loss or potential loss of the third barTier -fuel clad, RCS and/or containment.(B) A precursor event or condition that, unmitigated, may lead to a loss of all three fission product barriers. Precursor events and conditions of this type include those that lead directly to core damage and loss of containment integrity.(C) A release of radioactive materials to the environment that could result in doses greater than an EPA PAG at or beyond the site boundary.(D) A HOSTILE ACTION resulting in the loss of key safety functions (reactivity control, core cooling/RPV water level or RCS heat removal) or damage to spent fuel.3.1.5 4--4Risk-Informed Insights Emergency preparedness is a defense-in-depth measure that is independent of the assessed risk from any particular accident sequence; however, the development of an effective emergency classification scheme can benefit from a review of risk-based assessment results. To that end, the development and assignment of certain ICs and EALs also considered insights from several site-specific probabilistic safety assessments (PSA -also known as probabilistic risk assessment, PRA). Some generic insights from this review included: 1. Accident sequences involving a prolonged loss of all AC power are significant contributors to core damage frequency at many P.eS...iZ.d Water- Reaetai.. (PA1R) and B .iling Water Reaet.r.. (B...).For this reason, a loss of all AC power for greater than 15 minutes, with the plant at or above Hot Shutdown, was assigned an ECL of Site Area Emergency. SITE AREA EMERGENCY. Precursor events to a loss of all AC power were also included as an U:nSumal Event UNUSUAL EVENT and an A4e4-ALERT. A station blackout coping analyses performed in response to 10 CFR § 50.63 and Regulatory Guide 1.155, Station Blackout, may be used to determine a time-based criterion to demarcaie between a 8i4e Area Emer'geneySlTE AREA EMERGENCY and a General EmergencyGENERAL EMERGENCY. The time dimension is critical to a properly anticipatory emergency declaration since the goal is to maximize the time available for State and local officials to develop and implement offsite protective actions. STP is an Alternate AC plant and a Station Blackout battery copying analysis is not required.Nonetheless. a 125 VDC Battery Four Hour Coping Analysis was conducted and provides a basis for the timc-based escalation path from a SITE AREA EMERGENCY to a GENERAL EMERGENCY.

2. For severe core damage events, uncertainties exist in phenomena important to accident progressions leading to containment failure. Because of these uncertainties, predicting the status of containment integrity may be difficult under severe accident conditions.

This is why maintaining containment integrity alone following sequences leading to severe core damage is an insufficient basis for not escalating to a Gener-al EmergencyGENERAL EMERGENCY.

3. PSAs indicated that leading contributors to latent fatalities were sequences involving a containment bypass, a large Loss of Coolant Accident (LOCA) with early containment failure, a Station Blackout lasting longer than the site specific c .ping period four hours, and a reactor coolant pump seal failure.101 Page The generic EAL methodology needs to be sufficiently rigorous to address these sequences in a timely fashion.3.2 3-TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS The NEI 99 0! STPEGS methodology makes use of symptom-based, barrier-based and event-based ICs and EALs. Each type is discussed below.Symptomn-based ICs and EALs are parameters or conditions that are measurable over some range using plant instrumentation (e.g., core temperature, reactor coolant level, radiological effluent, etc.). When one or more of these parameters or conditions are off-normal, reactor operators will implement procedures to identify the probable cause(s) and take corrective action.Fission product barrier-based ICs and EALs are the subset of symptom-based EALs that refer specifically to the level of challenge to the principal barriers against the release of radioactive material from the reactor core to the environment.

These barriers are the fuel cladding, the reactor coolant system pressure boundary, and the containment. The barrier-based ICs and EALs consider the level of challenge to each individual barrier -potentially lost and lost -and the total number of barriers under challenge. Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. These include the failure of an automatic reactor scram/trip to shut down the reactor, natural phenomena (e.g., an earthquake), or man-made hazards such as a toxic gas release.3.3 NSSS DESIGN DIFFERENCE.S M--STPEGS DESIGN CONSIDERATIONS The NE! 99 01 emcrgeney cl-assification Scemlffe accountS fcr the design difefcenees between PWIRs and BWA's by speeifý'ing EALs uinique to each type of Nuclear Steam Supply System (NSSS). There are also significant design differencs aogPRNSs ther~efor-e, guidance is provided to aid in the development of EALs appropr~iate to different PAIR NSSS ty'pes. Where nlecessar~y, develcpment guidance also addresssuiu considerations for advanced non passive reacter designs sucih as the Advanced Beiling Water- Reactor (ABWR).the Advanced Priessuriz-ed-Water- Reactor- (APWIR) and the Evolutionarfy Powver RaeatorF (EP121).Developer~s will need to) consider the rek-levant aspects of their: plant's design and operating character-istics when onverting, the ggeneric guidance of this documzent into a site specific classification scheme. The goalist mainitain as much fidelilty as possible to the intent of generici this end, developers of a schemie for _An -advanced non passive reactor- may need to add, modify, or delete some infen~atiin. contained in this document; these changes will be r-eviewed ýfo acceptability by the NRC as part of the scheme approval process.The guidance in NEI 99 01 is not applicable to advanced passive light -ater reacto design.AEmrec Classification Sc~heme for this ty'pe Of planlt shoul1d be developed in arceeo Dcevelopmc;ew of E gcyieonLeve!s, Adwiacd Pas.qiivc Ligh Ut !fac constraints imposed by the plant design and ope.ating char'acteristics. To v'41 ,,;1. 07 0"1, Alt.. ed^.1ieg....4".... PeocttefslCs anld EALs within the...... ......... ..I.. .IlIIP agRe 3.3 The South Texas Project Electrical Generating Station (STPEGS) is composed of two units, each having an identical pressurized water reactor (PWR) Nuclear Steam Supply System (NSSS) and turbine generator (TG).The NSSS is a Westinghouse Electric Corporation four-loop PWR. High-pressure light water serves as the coolant. neutron moderator. reflector. and solvent for the neutron absorber. The Reactor Coolant System (RCS), comprised of four parallel loops (each with a RCP and a steam generator [SG]), is used to transfer the heat generated in the core to the SGs using RCPs to circulate the water. RCS pressure is maintained by means of a pressurizer attached to the hot leg of one of the loops. The RCS is designed to circulate borated demineralized water at temperatures. pressures and flow rates consistent with the design thermal and hydraulic performance of the NSSS.The Reactor Coolant Pressure Boundary Leak Detection System consists of temperature, level, humidity, and radioactivity sensors with associated instrumentation and alarms. Small leaks are detected by temperature and level changes of systems, increasing sump levels, and humidity and radioactivity concentration changes inside the Containment. Large leaks are detected by changes in reactor coolant inventory, changes in flow rates in process lines and changes in sump level.Emergency Core Cooling System consists of three independent trains. each one capable of providing 100 percent of the required flow to the core in the unlikely event of a LOCA. Each train consists of one high-head safety iniection pump and one low-head safety injection pump. Heat is removed from the system during recirculation by the residual heat removal heat exchangzer (low-head pump only). The piping and valving associated with each of the three subsystems are identical. In the event of a steam pipe rupture. the ECCS provides adequate shutdown capability. The Reactor Containment is a post-tensioned concrete cylinder with a steel liner plate, hemispherical top, and flat bottom. This structure provides a virtually leaktight barrier to prevent escape of fission products to the environment in the unlikely event of a loss of coolant accident (LOCA).--ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION 3.4 The scheme's generic information is organized by Recognition Category in the following order.* A-R- Abnormal Radiation Levels / Radiological Effluent -Section 6* C -Cold Shutdown / Refueling System Malfunction -Section 7* E -Independent Spent Fuel Storage Installation (ISFSI) -Section 8* F -Fission Product Barrier -Section 9* H -Hazards and Other Conditions Affecting Plant Safety -Section 10 S S -System Malfunction -Section II*PD Permaneiiflv Defuceled statben Atniefdix C 12 P age Each Recognition Category section contains a matrix showing the ICs and their associated emeigeney classifiation lcvcls.EMERGENCY CLASSIFICATION LEVELS. The following information and guidance is provided for each IC:* ECL -the assigned emergency classification level for the IC.* Initiating Condition -provides a sunmmary description of the emergency event or condition.

• Operating Mode Applicability

-Lists the modes during which the IC and associated EAL(s) are applicable (i.e., are to be used to classify events or conditions).

• Emergency Action Level(s) -Provides eXm'nples off cporS and indications that are considered to meet the intent of the IC.Devclepefs should add... .a.. e.ample EAL. if the generi. appr.a.h to the development .an example EAL cannot be used (e.g., an assumed instrumentation range is net avAilable at the plant), the developer atempt to specify an alterniate means for- identify'ing entry inte the 1G.For Recognition Category F, the fission product barrier thresholds are presented in tables BWRs alid PWRs, and arranged by fission product barrier and the degree of barrier challenge (i.e., potential loss or loss). This presentation method shows the synergism among the thresholds, and supports accurate assessments.

Basis -Provides background information that explains the intent and application of the IC and EALs. In some cases, the basis also includes relevant source information and references. incelude clarificaitions, r-efercenccs, e~amples, instructions for- calculations, etc. DeVeloper notes should not be inechded On the site's emer-gency classification scheme basis doeumcnt. Develapers may clect to ineludc infor-motioni resulting from n developer note nction in a basis scetion.EGAL ASSIgnment Attributes Located within the DcvclOpfr Notes scction, specifics the attr-ibute used feor assi-anina-the WC to it given ECL.--IC AND EAL MODE APPLICABILITY Q-3.35 The NEI 99 01 STPEGS emergency classification scheme was developed recognizing that the applicability of ICs and EALs will vary with plant mode. For example, some symptom-based ICs and EALs can be assessed only during the power operations, startup, or hot standby/shutdown modes of operation when all fission product barriers are in place, and plant instrumentation and safety systems are fully operational. In the cold shutdown and refueling modes, different symptom-based ICs and EALs will come into play to reflect the opening of systems for routine maintenance, the unavailability of some safety system components and the use of alternate instrumentation. The following table shows which Recognition Categories are applicable in each plant mode. The ICs and EALs for a given Recognition Category are applicable in the indicated modes.13 Page MODE OF APPLICABILITY MATRIX Recognition Cate~orv Mode AR C E F H P-1D S Power Operations X X X X X Startup X X X X X Hot Standby X X X X X Hot Shutdown X X X X X Cold Shutdown X X X X Refueling X X X X Defueled X X X X D'l-fulleffly _ _ _ _-STPEGS Operating Modes Mode Description Criteria (Rx Power excludes decay heat)1 Power Operations Reactor Power > 5%, Keff> 0.99 T Avi_> 350'F 2 Startup Reactor Power < 5%, Keff> 0.99 T Avg > 3507F Hot Standby Reactor Power 0% Keff< 0.99 T Avg > 350'F Hot Shutdown Reactor Power 0% Keff< 0.99 350 0 F > T Avg > 200°F Cold Shutdowvn Reactor Power 0% Keff< 0.99 T Avgy < 200°F 6 Rfuelina Reactor Power 0% Keff <0.95 T Avg < 140'F Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.Defueled All fuel removed from the reactor vessel (i.e., full core offload during refuel or extended outage)Developers wiAl need to intorporate the mode .riteria fram unit speeific Tehinical Speeifieations into their emerenc laifi n ..h.me. in addition, the sch.me must .also .icl.de the following miade deS s5pecific te N Hl 99 01: Deftieled (Noflre}7 All~ fuel1 rprn'v'd ffa il,,o m-ei v 1: -lr__ll I ý- I _ ._ý!.e --- -..ttktt tuitifttt ret~tlfix OF or dtrKiEWO outage).14 Page

4. SITE-SPECIFIC SCHEME DEVELOPMENT GUIDANCE 4 STPEGS SCHEME DEVELOPMENT effiegefe n Iasification scheme. COHn ptuall)' the appi-each diScussed here mirror~s the approeach Elsed to prepare emetrgency' operzatinig procedures generic material prepared by reactor: vendor- owners goups Is cneritted by each nucelear:

poer'e plant intot site specifie emier-gency cperatiing proeedures. Likewise, the emferg-encyý 6iassination Snemcl OeVeioper' WHI USe tile genieri. gUimoanee 1n NEI 99W tO prepare a site speecmcl emerg-ency classifiationi scheme and the associated basis docuiment. it is important that the NEI 99 01 emer~gency classification scheme be implemented as an integrated packlage.Selected uise of por-tions of this guidance is stron~gly discour-aged as it will lead to an inconsistent ori incomplete emerg ..ency classification scheme that will likely not .eceive the necessary reg.lat.y appreval-a.1 GENERAL IMPLEMENTATION GU 1DANCE 4.1 GENERAL DEVELOPMENT PROCESS The guidance in NEI 99 01 is not intended to be applied to plants "as is"; however, developers should attempt to keep their site specific schemes as close to the generic guidance as possible. The go.-al is to meet4 the intent of th geerc nittiat-in-g Coniditions (IC-s) and Emer:.gency Actioni Levels (EA~s) withini the conlteEt of site speeific chara.cteA.....ri.stices locale, plant design. ope.ating terminology, etc. Meeting this goal will r.esult in a shorter and less cumffberso~me NRC review and approeval process, closer alignment with the schemes of other nuclear power plant sites and befter pesitiening to adopt future industfý' w..ide scheme enhancements. The STPEGS ICs and EALs were developed to When. properly' developed. the ..s and EALs should be unambiguous and readily assessable because both serve specific purposes. As discu-ssed in Section 3, the generic 9 .......R. 4;C ...... I _ ..e a..... o guidance-. inldssandd example EALs. it is the intent of thiis guidancee th~at both be included in site specific documents as eac. serves a specific purpose. The IC is the fundamental event or condition requiring a declaration. The EAL(s) is the pre-determined threshold that defines when the IC is met. To this end, the STPEGS ICs and EALs were developed with input from key stakeholders such as Operations, Training, Health Physics. and Engineering. STPEGS specific indications, parameters and values were consistent with licensing basis documents, plant procedures, training, calculations. and drawings f some fetr of the p location or design is not .ompatible with a generic IC or EA, efforts should be mlade to idntif- an alternate IC AL.If an IC omv EAL includes an explicit reference to a mode dependent technical specification limit that is not applicable to the plant, then that WC and/or- EAL need noet be inceluded in the sit spegeshemze. ini thsess dev~elopers must pr-ovide adequate documentation to justify, why the IC and/or EAL wyere no0t incorporated (i.e., sufficient detail to allew~ a thir-d party' to uniderstanid the deeisi, i ~net to ineorpor-ate the generic guidance). Useful acronyms and abbreviations associated with the NEI 99 01 STPEGS emergency classification scheme are presented in Appendix A, Acronyms and Abbreviations. Those specific to STPEGS were included to be consistent with site terminology. site procedure, and training.Site specific entries may be added if necssary.Many words or terms used in the NE! 99 0- STPEGS emergency classification scheme have scheme-specific definitions. These words and terms are identified by being set in all capital letters (i.e., ALL CAPS). The definitions are presented in Appendix B, Definitions. 15 Page B-elow. Are Pex-amples of acceptable modifications to the generici guidance. Týhese may be incor-por-ated depending upon Site develOper anid user pr-efcrcncceS. The W-s w~ithin a Rec-ognition Category, may be placed iin reverse cr-der for presentation purFpEseS (e.g., start with a Gener-al Emiergency at the left/tep of a user aid, followed by Site Area Emer-gencey, Alert and NOUE)+.The Initiating Condition numbe-ing may be chianged.The First letter of a Rcco,,..gnition Category, designation may be changed, as follows, previded the ehange is car-ied through for all of the associated 1G identifiers&. -R may be used in lieu of A-N4 may' be used 6in lieu of9 For e .ample, the Abnorm.al Radiation Levels / Radioloagial Effluent category designator "A" (for Abnormal)may be c -ianged to "R" (for Radiation). This means that the associated

Cs would be chan.ged to RU ... RU2, The .... and PE4ALs from RWecgnition Categories S am d C may, be incopo rated inte a eammon presentation method (e.g., ane table) privided that all related nmtes and mide applicability arequirements ae maintained.

The ICs and EALs fO Eerigency Diret Aijudgmiient and security related evsents may be placed under separate Recognition Categoraies. The terms EAL and thheshold may be used intfacliangeablt. The material in the Developer Notes section is included to assist developers with c heafig corrfcat iG alid EA Sateented. ThiS mateial is not required to be in the final energency classification scheme basis document..

4.2 CRITICAL

CHARACTERISTICS As discussed abov~e, developers are encouraged to keep their site specific schemes as close to the genieric guidancee as possible. When crafting the scheme, developer: should satisPv themiselvesSTPE.GS ensured that certain critical characteristics haverbeen wereimet. These critical characteristics are listed below.The ICs, EALs, Operating Mode Applicability criteria, Notes and Basis information are consistent with industry guidance; while the actual wording may be different from NEI 99-01 Revision 6, the classification intent is maintained. With respect to Recognition Category F., a-the STPEGS scheme included a sei espeeille sc-hemel muIst inlu~kde some type3 of uLser-aid to facilitate timely and accurate classification of fission product barrier losses and/or potential losses. The user-aid logic is-nmust-be consistent with the classification logic presented in Section 9.*EAL statemnents use obljective criteria and observable values.* Cs.. EALs, Operating Mode Applicability and Note statements and formatting consider human factors and are user-friendly.

• The scheme facilitates upgrading and downgrading of the emergency classification where necessary.
• The scheme facilitates classification of multiple concurrent events or conditions.

16 Pa P e

4.3 INSTRUMENTATION

USED FOR EALS lnStrum1entation referenced in EAL statements shudinelude that described in the. Pemergency plan section which addr-esses 10 CFR 50.47(b)(8) and (9) and/Or Chapter 7 of the FSAR. !nstrumentation utsed fer E;A~s need niet be safety related, addr-essed by a Techinical Specificationi OF ODCNM/TS controal r-equirement.nr HOwF e fro"--t 0M.an eme'rgency pewer source; h.weve.., EAL developers .. ould stri to incorporate STPEGS incorporated instrumentation that is reliable and routinely maintained in accordance with site programs and procedures. Alarms referenced in EAL statements should-be are those that are the most operationally significant for the described event or condition. Scheme developers should ensure that specified values used as EAL setpoints are within the calibrated range of the referenced instrumentation, and consider any automatic instrumentation functions that may impact accurate EAL assessment. In addition., EAL setpoint values sheum4-do not use terms such as "off-scale low" or "off-scale high" since that type of reading may not be readily differentiated from an instrument failure. Findkhgs-and violationis related to E ins. t rum....entationl may be locat-ad on the NR. website. If instrumentation failures n. Fit i 0 I .r A I c 00011, ot1A,it* t-n,, (1, ---oc --,1 tn, , aI~anirninco-~, f n n r implementation may be used as described in plant procedures.

L- 4.4 PRESENTATION OF SCHEME INFOPAMATION TO USERS The US Nuclear 'egulatoi' Comzm..ission. (NRC=) e..pects licensees to establish and maintain the capability to assess, eclassify and declare anl em .ergenc.y c.ndition prmptly within 15 minutes aer the availabilit of inidicationis to plant operators that an emergeney, action level has been, or- may be, eEceeded.

When wiin an emergency classification procere and creating r.elated user aids., the developer must determine the presentation maethod(s) that best supports the enid users by, fac-ilitating accurfate and timely c mergency classification. To) this end, developer-s shouild conisider the following, points.The first users ofan em.er.gency classification! afe the operators in the Contro.l Rom. During the ae .lassifcati time period, theym ay... have responsibility. toWperfor oppther crFiticea tasks, and 'ill likel, have minimal assistance in m~aking, a classification assessmient. Asan emer-gency situation evolves, members ofthe Control Room staffare likely to be the first personnel to no~tice a change i pAn cn itins. They can assess the chianged conditions and, when w 1 am~anted, r-ecommenld a di:fferent e-m-l ier, en cassific-ation level to) the Teehnie-al Support Center (T-SC) and/or Em3ergency) Oprtin Faeility (EOF). Emergency Directors in the T-SC- and/or EO.F will have moree appetlunity to focus on making an emergencey classification, and will probably have advisors from Operations available to help thema.Emergency classification sc-heme inform1ation for end- user-s should be presenlted in a Manner with which licenfsead operators are mROSt comfort1-14able. Developers V.ill neead to wok losely With repr-esentatives from the Operations and Operationis Triainling Departmnents to develop readily usable and easily understood classificationi tools (e.g-., a proced~ire and relate uiser aids.). if necessary, an altern-atea mfethod for- prsetigemergency classific-ation schemfe information m~a" be developed for- use by Emiergency Directors anidor Offsite Response Organizationi pe!-efflel ..... ... ........ ................. .. ---........ 1....7 P .a. ...17 1PagP-e r wallboard is a.. aeptable presentation ethOd p. Videeds that it e antains all the iinfo-,.ati.n niecessary to make a eeilfeet emnergene)y classification. This informiatien inceludes the !Gs, Operatinig Mode Applieability, criteria, EALs and Notes. Notes may be kept w.'ith each applic-able EAL or moved to a common area and referenceed; a r-efer-ence to a Note is acceptable as long as the inform4:ationi is adequiately captur-ed on the wallboard and Pointed to by eacl applicable EAL,. Basis informa.io.nee not, be included an a A .allboad but it should be readily available to em.ergency, classification decision m.ak.ers.in some cases, it may be advantageous to develop two wallba-ds one for- se during power oper1ations, startup and hot conditions, and another- for cold shutdoAwnr -And rek-fueling conjdition+s. Alternlative preetto ethods for91 the Recognition Categor-y F !Gs and fission pr-oduct bapr-rierf threSholds ar-e acceptable and include fieow charts, block diagramns, and celieklist type tables. Developer-s must ensurfe that thle site specific method- ad-dre-sses all possible threshld cmbnaiosndclsic atio otcoeels Shown in the BAIR or PAIR EAL fission product barrier tables. The NRC staff considers the presentation m:ethod of the Recognition Category, F information to be an impoitanit user aid and may' request a change to a particutlar pro)posed mehoL4d if. among other- reasonis, the chiange is necessary to promote ceonsi-stency acroass the industr-y. A Fgoeusinegratonof IC and;H E2.AL reaferencees into plant oper-ating proeedurFes isntreemmenueu. TisL approach would gr-eatly inresethe admfinlistrative cOntrols anid workload fer mnainitaining proeedures.On-the other hanld, performance ch-allenges mayf occur if recognitionl ofFmeeting -An IC- Orf EL4 i-S bazed SelelY, On tile ffemom~y of a licensed operator or an Emergency Direector , especially during periods of high stress.Dev~elopers should eonsider placinig approepriate v.isual cuies ( e.g., a step, note. caut~ion. etc.) in plant pim eedui-e alerting the reoadder/uIster toe con;;s-Ul th~e site. emergency classific-ation procedur-e. Visual cues could be placed in emergency) operaltinig procedures, abnormlal operating procedures. alarm responise procedures, anid norma operating procedures that apply to cold shutdown and r-efueling moades. As an exiample, a step, no~te- or cauMtion could be placed at the beging of n CSlea abnorma o per-ating proeedurFe that rmnsthe rea;der that an emergency) classifiation assessmenit shouild be perbformed. 4 6 -BASIS-DOCJLANT A basis docueknet is an integral part ofan emer-gency classification scemine. The material in this docuent~fe supports proper- emergency cl-assific-ationA deiio aking. by' prov'iding, iniforming backigro-und and development in4;fonnationf in a readily accessible formiat. it can be r-eferrFed to in training situations and wheni making ani actulal emergency classification, if necessary. The docue-ment is also; useful for establishing configuration management cntrols for EP related equipmient and exiplaining.. aneergeny classification to off-site authorities. The content offthe basis docuiment should include., at a mninimuma, the following: A site specific Mode Applicability NMatrix anld deseription of oper-ating modes, similar to that pr-esenited in section 3.5.A discussion of the emnergency classification and declaration proceess reflecting the material presented in Sectiont-5. Th-is matrilmay be edited -as needed- to align withi site specific emnerg@eny plan and implementing proc-edure Eachi initiating Condition along with the associated EALs or- fission pr-oducet barrier thr-esholds, Operating Mode Applicability, Notes and Basis inforemation.o 18 1 P aa e stin',g of acronvm~s A dfind te....~rm similar t,~ tha ......et ad i:n defdici A -ndl B respectivel" Thi.material may be e~ditted as ineeteded to ali-gn with site speeific chiaracteristics.- Anyi) site specific baekgrounifd or technical appenidices that the developer-s believe would be usefuil in explaining or using elements of the emergency classification scheme.AO Basis section should noat contain that co-ld mIdIfy the meaning or intent of the associated C r I-, 1 1, in the B.asis should nly .la.ify.. and infor.m decision making fob*- an eni.g.n.y .lassification. Basis informatie~n shoul:d be rcadi!3' available te be i'cfercne',d, ifiieeessar-y, by the Emeirgeney3 Dir-ector. Fer exiample, a copy of the basis document could be maintained in the appropriate emergency r-esponse facilities.

4.4 REFERENCES

TO STPEGS AOPS AND EOPS As reflected in the generic guidance, Some of the criteria/values used in several EALs and fission product barrier thresholds may-be were drawn from a p!a;44t'sSTPEGS AOPs and EOPs. This approach i-was intended to maintain good alignment between operational diagnoses and emergency classification assessments. e-elepei-s shakild v'e'i4-4 th-at STPEGS verified the appropriate administrative controls are in place to ensure that a subsequent change to an AOP or EOP is screened to determine if an evaluation pursuant to 10 CFR 50.54(q) is required.s GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS-i5.1_GENERAL CONSIDERATIONS When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basis information. In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL.NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-0 1, Interim Staff Guidance, Emergency Planning for Nuclear Power Plants.All emergency classification assessments should be based upon *4a-aVALID indications, reports or conditions. A v#ald VALID indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, validation could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. The validation of indications should be completed in a manner that supports timely emergency declaration. 191Page For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that 1) the activity proceeds as planned and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 § CFR 50.72.The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.); the EAL and/or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift). While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99 01 This scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated into the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.5.2 CLASSIFICATION METHODOLOGY To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL(s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, then the IC is considered met and the associated ECL is declared in accordance with plant procedures. When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock." For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01.

5.3 CLASSIFICATION

OF MULTIPLE EVENTS AND CONDITIONS When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:--- 201Page

• If an A4er4ALERT EAL and a Site Area EmergencSITE AREA EMERGENCY EAL are met, whether at one unit or at two different units, a Site ea Emergency'SITE AREA EMERGENCY should be declared.There is no "additive" effect friom multiple EALs meeting the same ECL. For example:* If two A4eAALERT EALs are met, whether at one unit or at two different units, an A-ei4ALERT should be declared.Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events.5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable.

If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher.5.5 CLASSIFICATION OF IMMINENT CONDITIONS Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all cmergency cla.sificatin 4le4esEMERGENCY CLASSIFICATION LEVELS, this approach is particularly important at the higher em.ergency ela.sifieatien leve EMERGENCY CLASSIFICATION LEVELS since it provides additional time for implementation of protective measures.21 IPaae

5.6 EMERGENCY

CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated. The following approach to downgrading or terminating an ECL is recommended. ECL Action When Condition No Longer Exists Unusual Event UNUSUAL EVENT Terminate the emergency in accordance with plant procedures. A4e4ALERT Downgrade or terminate the emergency in accordance with plant procedures. Site Area ...... egneySITE AREA Downgrade or terminate the emergency in EMERGENCY with no long-term plant accordance with plant procedures. damage Site Area AREA Terminate the emergency and enter EMERGENCY with long-term plant recovery in accordance with plant damage procedures. Genera! E:mergencyGENERAL Terminate the emergency and enter EMERGENCY recovery in accordance with plant procedures. As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02.2,65.7CLASSIFICATION OF SHORT-LIVED EVENTS As discussed in Section 3.2, event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus., over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include a failure of the reactor protection system to automatically scrain/trip the reactor followed by a successful manual scram/trip or an earthquake.

5.8 CLASSIFICATION

OF TRANSIENT CONDITIONS Many of the ICs and/or EALs contained in this document employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions. 221 P a e EAL momentarily met during expected plant response -In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures. EAL momentarily met but the condition is corrected prior to an emergency declaration -If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example.An ATWS occurs and the auxiliary feedwater system fails to automatically start. Steam generator levels rapidly deerease lower and the plant enters an inadequate RCS heat removal condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts the auxiliary feedwater system in accordance with an EOP step and clears the inadequate RCS heat removal condition prior to an emergency declaration, then the classification should be based on the ATWS only.It is important to stress that the 15-minute emergency classification assessment period is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event; emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations where an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions. 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements. 5.10 RETRACTION OF AN EMERGENCY DECLARATION Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-l 022.23 1 Page 6 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/EALS Table AR-I: Recognition Category "AR" Initiating Condition Matrix UNUSUAL EVENT A-U4RUI Release of gaseous or liquid radioactivity greater than 2 times the (sie specific effluent r-elease controlling doeument)ODCM limits for 60 minutes or longer.Op. Modes: All A-IRU2 UNPLANNED loss of water level above irradiated fuel.Op. Modes: All ALERT SITE AREA EMERGENCY GENERAL EMERGENCY AA4RA1 Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrenl TEDE or 50 mrem thyfei4THYROID CDE.Op. Modes: All AA-RA2 Significant lowering of water level above, or damage to, irradiated fuel.Op. Alodes: All A-A-3RA3 Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.Op. AM'odes: All ASIRSI Release of gaseous radioactivity resulting ill offsite dose greater than 100 mrem TEDE or 500 mrem thyro4iTHYROID CDE.Op. Modes: All A2RS_2 Spent fuel pool level at 40'-4" or lower (site speci: Ae.el 3 dese~iptian4 Op. M~odes: All A-4RG_ Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyiredTHYROID CDE.Op. Modes: All A-GRG.2 Spent fuel pool level cannot be restored to at least 40'-4'- (site specific Level 3 des....p.iefi. for 60 minutes or longer.Op. Modes: All 241 P a e AUMRU1 ECL: Netiieati AP of Unusual Event UNUSUAL EVENT Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the (site speeifie cffluent release controlling document) ODCM limits for 60 minutes or longer.Operating Mode Applicability: A4IALL FefxF-Ape-Emergency Action Levels: (1 or 2 or 3)Notes:* The Emergency Director should declare the Unusual Event UNUSUAL EVENT promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer alTid-VALID for classification purposes.(1) Reading on ANY of the following effliwiet-radiation monitor greater than 2 times the (site specific efflueint release contro.lling, document) limits shown the values listed in Table RI column "UE" shew~for 60 minutes or longer: (site specific 1 I ....4 monitor list and4 tl~rcs118dt Values ean-espn~ndan to 2 times the eentiRilln-lsmn limits)Release Point Unit Vent Main Steam Lines Monitor RT-801 OB RT-8046 thni 8049-1-1.50 E+Table RI: Effluent Monitors GE SAE ALERT 08 ttCi/sec 1.50 E+07 utCi/sec 1.50 E+06 uCi/sec 02 OCi/cni 3 4.00 E+01 gCi/cm 3 4.00_E+00 pCi/cm 3 UE 1.40 E+05 tCi/sec 5.00 E-02 LCi/C rn 3 4.00 E+(2)Reading on .ANY effluent r-adiation monitor. g.eater than 2 times the alar setpoint established by a.....R on gaseous effluent radiation monitor RT-8010B greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.(3) Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the (site specific, efent ..elea:. document)ODCM. limits for 60 minutes or longer.Basis: 25IPage This IC addresses a potential Eee-easelowering in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.Nuclear pcwe" plants STPEGS incorporated design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the enviromnent is indicative of degradation in these features and/or controls.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer vali4-VALID for classification purposes.Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.EAL #1- This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.EAL #2- This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).EAL #3- This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).Escalation of the emergency classification level would be via ICA-A4RA J.RUI: EAL-1 Selection Basis The Unit Vent and Main Steam Line monitor readings were included in this EAL because they give instantaneous indications of a monitored gaseous release exceeding twice the ODCM limits. Normal gaseous effluents are due to planned RCB purges and monitored by the Unit Vent. The Main Steam Line monitor readings were included because they correspond to a concentration that Would result in a release rate of twice the ODCM limits if there were a release via the Power Operated Relief Valves (PORVs) or Safety Relief Valves. A release from the PORVs or Safety Relief Valves is not a normal effluent pathway but meets the intent of the EAL.The Unit Vent and Main Steam Line release values are based on Calculation No. STPNOCOI3-CALC-002, Rev.26 1 P a R e RUl: EAL-2. 3 Selection Basis For EAL-2, there are two effluent radiation monitors, RT-8038 (liquid) and RT-8010B (gaseous). however only RT-8010B was included. The alarm setpoint for the gaseous effluent radiation monitor RT-8010B is set at the ODCM limits. An indication of two times the alarm setpoint (two times the ODCM limit) would allow operators time to secure the release prior to meeting this EAL. The liquid effluent radiation monitor RT-8038 was not included in EAL-2 because the activity in liquid dischariges is normally the several orders of magnitude lower than the ODCM limits. In order to alert personnel to significant changes in the liquid effluent activity, the alarm setpoint for RT-8038 is normally set several orders of magnitude below the ODCM limits. Setting the alarm setpoint for RT-8038 at the ODCM limit would remove this capability and violate the intent of the EAL.For EAL-3, sample analysis could be used as a backup for the effluent monitor indications.

REFERENCES:

I. Calculation No: STPNOC0l3-CALC-002 Rev. 2, Radiological Release Thresholds for Emergency Action Levels 2. Offsite Dose Calculation Manual (ODCM), Rev. 17. Part B3.0 to B4.9 3. UFSAR, Rev. 14. Section 11.5.2.3.3 and 11.5.2.5.3 (monitor descriptions)

4. UFSAR, Rev. 14. Section 11.5.2.4.4 (liquid waste processing monitor)271 P ag e Dwlpei- Netes: The "site sp"ific e1-fifent felease 1-fntr.lling is the Radiol(o)gi÷al Effluent Te"hnical SpeTifications

(,T-S) or, for plants that have implemented Generic L"tter. 89 014, the ffsitD Pose Calculation Man1ual ThL11. ese Enocuments implement r-eguipuions reiaico to eT-luent controls, c.g.. Iu Q~ Fart -,w ang Lu v -ltt Part 50, I). As appropriate, the RETS ..r COGrC- methodology should be .sed foF establishing the monitor thlresholds for this WC.Listed monitors should include the effluent mnitors descri-bed in the RETS or ODCM.DeVelopers may also ..nsid.. ,inluding, installed Monitors assoiat.d with other potential effluent pathways that ae no.t described in the RETS or OD.M..6. if.inluded, EAL values fo. these nit1orS Should be d.tes..ined using the most applicable dosc,trelcasc iimis p-rs-nted in the RETS orF ODCM. it is rcoegnizced that a ealcuiatcd EAL value may be below wh.at the mo.nitor an Fad; in that case, the mniitor does not need to be included in the list. Also, soen moiosmy notý be governed by Technicial Speciflcationis or other licenise related r-elated requiremenits; ther-efoare, it is impor'tant that the -a-s-soiated EAL and basis sete learly identify any limaitations onH the Else or availability offthese moenitor-s. Some sites m.ay 'find it advantageous to address gaseous and liquid releases with separate EALs.Radiation monitor readings should refect values that correspond to a radioelgical release exc.eeding 2 times release control limiit. The eontrollinig docuimenit typically describes methodologies fo9r deterinRing~l effluent radiationR monitor setpoints; these methodologies should be used to determine EAL .alues. in cases where a methodologyý is not adequately defined, develepe~s hould determ~ine valuies consistent with effluent conitrol regulations (e.g., 10 CF.R Pa .20 -And- 10 CFR Prt 50 .ppendix ) and related guidance.Fr-f EAL #t2 Values in this EAL should be 2 times the setpoint established by the radioactivity dischiarge per-mit to w.arn efa release that is not in compliance with the specified limits. indexing the value in this manner enstres conisistency) between the EAL and the setpoinit established by a specific dischar-ge Pe!-i!Developers shouild researchi radiation monitor design documents of, other- information sources to ensure that 1)the EAL value being consider-ed is within the usable r-esponse and display range of the instrument, and 2) there arc no au.tomati features that may render the monitor reading invalid (e.g., an auto purge feature triggered at a pariciular indicationi level).it is r-ecognized that the condition descr-ibed by this IC- may result in a radiological effluent value beya'nd the operating or display range of the installed effluent monitor. .in those cases., EAL values shoud be with a m .argin .sufficien't to en.sur.e that an accurate onfitor reading is available. For ex'ample. an. EAL mo.nitor Feading minght be sei at 900% to 95,% of the highest accurate monitor i-adn. This prov~ision notwithstanding, if the estimated/calcuilated monitor r-eading is greater than approxnnately 110% of the highest accurate monitor ,,reain, then developers may cheose not to inc.lude the mOnito as an indication and identify an alteinate EAL tlheshold. indications fromi a real time dose pricejetion system are not inlu~tded ini the genieric E.Ats. Many, liceense do. not ha~ve this capability. For those that do., the capability may not be within the scope Ef Whipat Teehnical Specifications. A licensee may r-equest to include an EAL uising, real timie dose projection system r~esults;approv~al will be conisidered an a case by case basis.28 1 P a g e llidieatiEHIS fr-OM a Perimleter mOnitering system ar~e not included iii the gencrii EALS. Many licensees do Hot have this capability. For- these that do, these monitor-s may not be controlled and maintained to the same le'vel as plant equipment, orf w~ithin the seope of the plant Technical Specifications. in addition, readings May be influenced by environm:eiital or- ethei factors. A licenfsee may r-equest to ineludc an EAL using a perimieter nienitoring~ sytm prvlWill be CEnIISidered en a case by ease basis.ECL AssigajnmentAttributes: 3.l1.l1.B 29 1 P a i4 e A2 RU2 ECL: Netifieation of Uiusual Ev'ent UNUSUAL EVENT Initiating Condition: UNPLANNED loss of water level above irradiated fuel.Operating Mode Applicability: -I4ALL EkI ample-Emergency Action Levels: (2) a. (site speDific level indidations). (1)~ a. UNPLANNED water level dron iii the REFUELING PATHWAY as indicated by AN-NAANY of the following:

• Visual Observation OR* Annunciator alarm on lampbox 22M02 Window F-5 "SFP WATER LVL HI/LO" OR* Spent fuel in the ICSA AND Annunciator alarm on lampbox 22M02 Window F-6 "SFP Trouble" AND Plant Computer point FCLC 1420 "REFLNG CAV LVL IN CNTMT" (ICSA Water Level HI/LO) is in alarm AND AND Din.t÷ /"a* ., ,- .,,.÷ Ef'i[ C' 1 A~f ";D1ZE, NITC2 A \1i r \irIX I Nl AT"l/q 11ir A I~l÷ I LIT/["IN ( o t..b. UNPLANNED rise in area radiation levels on ANY of the following radiation monitors.1. (site spe-ific list fCarea radiatien mo6nitols) 0RE-8055 (68' RCB) -Mode 5 or 6 only OR" RE-8099 (68' RCB) -Mode 5 or 6 only OR" RE-8090 (68' FHB)301 P ag e Basis: This IC addresses a deefease-lowering in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.A water level d-eerease-lowering will be primarily determined by indications from available level instrumentation.

Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations ble). A significant drop in the water level may also cause an inoaeaserise in the radiation levels of adjacent areas that can be detected by monitors in those locations, The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may ineFeaserise due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level.A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.Escalation of the emergency classifieation lec" EMERGENCY CLASSIFICATION LEVEL would be via IC AA-2RA2.RU2: EAL-1 Selection Basis Hi/Lo level sensors are located in the Spent Fuel Pool (LSHL 1401) and the RCB, In Containment Storage Area (ICSA) (LSHL 1420). If level in the Spent Fuel Pool rises or lowers by more than 6 inches above or below the normal water level of 66'-6" (UFSAR 9.1.2.1), the "SFP WATER LEVEL HI/LO" lampbox 22M02 window F-5 annunciator alarm is received in the Control Room (0POP09-AN-22M2. Annunciator Lampbox 22M02 Response Instructions). Although the ICSA has a Hi/LO level sensor, there is not an annunciator in the Control Room similar to the one for the Spent Fuel Pool. There is however, a "SFP TROUBLE" lampbox 22M02 window F-6 annunciator in the control room. One of the inputs to this alarm is FC-LSHL-1420, the ICSA HI/LO level sensor. Since no fuel is located in the ICSA in modes 1-4. this EAL only applies in modes 5 or 6.Area radiation monitors RE-8055 and RE-8099 are located are located in the RCB 68' elevation on the bioshield wall close to the refueling cavity. Area radiation monitor RE-8090 is located in the Fuel Handling Building on 68' Elevation near the Spent Fuel Pool.Expected radiation levels for a loss of water level can range from a few mR/hr to thousands of R/hr.For a drop of water level of approximately 14' (from 66'-6" to 51 '-10") with approximately 13' of water over the top of any array, the dose rate would be expected not to exceed 2.5 mR/hr, above background. This assumes 42 hours of decay with a full core off-load (section 9 of STP UFSAR).For a significant drop of water level that would still cover the arrays, the radiation levels could range from several hundred R/hr to over a thousand R/hr on and around the 68' elevation deck (table C-S NUREG CR/0649).311 Pa e

REFERENCES:

1. OPOP09-AN-22M2, Rev. 25, Annunciator Lampbox 22M02 Response Instructions F-5 and F-6 Window (level alamns)2. OPOP04-FC-0001.

Rev. 29, Loss of Spent Fuel Pool Level or Cooling (level alaims)3. Technical Specification, amendment 104 (Unit 1) and 91 (Unit 2). Section 5.6.2 (Design water level)4. UFSAR, Rev. 16. Section 9.1.2.1 (Dose rates)5. UFSAR, Rev. 16, Section 9.1.2.2 (Normal water level)6. NUJREG CR/0649 (Dose rates), reference only (not included in submittal)

7. Drawing 5R219F05028#1 Spent Fuel Pool Cooling and Cleanup System (level sensors)8. UFSAR, Rev. 15, table 12.3.4-1, Area Radiation Monitors Developer-Notes.: The "site specific lev~el indieatiens" are those inidications that may be used to moenitorf wAater level in the ya~el ff1ions;;

of the R-EFUELING PATHWAYA2. Specify the made applicability, of a pailiculaf indication if it is not avwailable in all moedes.The "site specific list of area radiation monitors"~ should contain toear-ea r-adiation monitor-s that would be expected to have incr-eased readings following a decrease in water- lev~el in the site specific REFUELING PcAsTHWJA'Y. in cases where a radiation monitor(s) is not available Orfu .ld not provide a useful indic.ation. conidratonshould be given to includinge alternate indications suchl as UNPLANNED changes in tank and/or stimplevels. Development Of the EALS shoulld- consid-er the -availability and limitations of mode dependent. Or other c-ontrolled but temperaly,, r-adiation mionitors. Speeif.' the moede applicability of a pai~ieulaf monflitor if it is not available in all modes.iý-Al Ac-cza.111 A , 321Pai4e AAMRA1 ECL: ,k!e4-ALERT Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrern TEDE oi 50 rnrem thyiad-THYROID CDE.Operating Mode Applicability: AIIALL xamfpe-Emergency Action Levels: (1 or 2 or 3 or 4)Notes:* The Emergency Director should declare the A-l.-t ALERT promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.0 If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer a4lid-VALID for classification purposes.* The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. (1) Reading on ANYANY of the following radiation monitors greater than the reading sho-wn values listed in Table RI column "ALERT" for 15 minutes or longer: (Site specific M...itor liSt and values)Table RI: Effluent Monitors Release Point Monitor GE SAE ALERT UE Unit Vent RT-8010B 1.50 E+08 tLCi/seC 1.50 E+07 LtCi/sec 1.50 E+06 LICi/sec 1.40 E+05 gtCi/sec Main Steam IT-8046 thru 4.00 E+02 uiCi/cn 3 4.00 E+01 ýLCi/Crfl 3 4.00 E+00 LICi/cm 3 5.00 E-02 [1Ci/cmn 3 Lines 8049 (2) Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyfeid-THYROID CDE at or beyond (site specific dose re.epter point) the SITE BOUNDARY.(3) Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyfaid-THYROID CDE at or beyond (site specifi e4--se i'eeepteF-pe4*nthe SITE BOUNDARY for one hour of exposure.(4) Field survey results indicate EITHER of the following at or beyond (site specific dose recepter poino the SITE BOUNDARY:* Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate thyreidTHYROID CDE greater than 50 mrem for one hour of inhalation. 331 P age Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Proteetive Action GuidesPROTECTIVE ACTION GUIDES -(PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significantuncontrolled release).Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem 4ph*eiTHYROID CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thy-ei4THYROID CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer a4,4diVALID for classification purposes.Escalation of the emergency AS-I-RS 1.C Iacztiteation 4eveI-EMERGENCY CLASSIFICATION LEVEL would be via IC RAI: EAL-1 Selection Basis The Unit Vent and Main Steam Line monitor readings were included in this EAL because they give instantaneous indications of a monitored gaseous release meeting the EAL threshold values of 10 mnrem TEDE or 50 mrem CDE THYROID at the SITE BOUNDARY. Gaseous releases from the plant are monitored by the Unit Vent. The Main Steam Line monitor readings correspond to a concentration that would result in a release rate meeting the EAL threshold values if there were a release via the Power Operated Relief Valves (PORVs) or Safety Relief Valves.The Unit Vent and Main Steam Line release values are based on Calculation No. STPNOC013-CALC-002. Rev. 2. The adiusted values used in this EAL were conservatively truncated by less than 1% of the calculated values to ensure they are readily assessable. RAI: EAL-2. 3. 4 Selection Basis N/A

REFERENCES:

1. Calculation No: STPNOCO13-CALC-002 Rev. 2.. Radiological Release Thresholds for Emergency Action Levels 2. UFSAR, Rev. 14, Section 11.5.2.3.3 and 11.5.2.5.3 (monitor descriptions)

3. UFSAR. Rev. 14, 11.5.2.4.4 (liquid waste processing monitor)De'clopes-Notes: While this WC may not be met absent challenges to One Or More produet barriers, it pravides classification diversity aind maN, be bisee to elassify events tha weule not r-eaefn Vie saie +-l-, based on pliant statuls or tne 341 P age fiSSionI product m-atrix, -Al]e For- many of the DR3A s anflalyz~ed in the UP dated Final Safety Analysis Report, the diSfrimfinlatOr VAIl not be0 thPe numfber- offission producet barriiers challenged, bult rather the amlOUnt Of rad ioactiyvity reeased to the environm~ent.

The EPA PAGS arc expressed in tearms of the suim of the effective dose equivalent (EDE) and the comm~itted effective dose equivalent (CEDE), or as thie thyroid coEmmitted dose, equivalent (C-DE). For the pur-ose of these IdEALs, the dose quantity total effective dose equiivalent (TEDE), as defined InQ1 C-FR § 20, is used in lieu ot".. .snum, of .E... and..CED........ The EPA PAG guidance p ,rovides fer the use of ad.ilt thr,, oid dose conversion factors; some states have decided to base ptetie ations on child th)yroid GDE. Nuclear powver plant IGsiEAtLs need to be consistent with the protec-ti've action mnethodologies employed by the States within their EPZs. The thyroid CDE dose utsed in the WC and EALs should be adjusted as necessary to aligni with State protectiv~e action decision making The "site specific monitor list and thr-eshold values" should be determiined with consideration of the followinigý" Selection of the appropriate installed gaseous and liquid effluent monitor's." The effluent !+oenitor readings should cofrespond to a dose of 10 mIwrc TEDE or. 50 thyroid IDE at the "site specific dose r-eceptor-point" (conisistent with the calculation methodology employed) foar onie hour Moio r4141 eadings will be calcullatead using a set of assumied meteorological data Or atmospheric dispersion! fac~tors; the d-ata Or factorFs seleRcted-for_ Use should be the- s-ameL aFS those employed to calcuilate the moitor4 readings Bfr WsC and AG I. Acceptable sour.es of*this infrm.ation include, but are not limited to, the RETS/ODCM anld v~alues used in the site's emiergency dose assessment methodology.-v" The calculation of monitor- readings wvill also requir-e use ofan assuimed r-elease isotopic mix; the selected mix should be the same as that employed to calculate Monitor readinigs fori W~s AS!I and AG 1. Acceptable sore f this; in;formation incjlude, but are-1 noAt limitead to, the RE2T-S/ODCN4 and vp-alues useLd in4 the. site'sLr em.r .e ...... A assessment methodology..Depending upon the methodology' used to calclate the EAL values, ther.e i..a:. be ov..erlap of some values between different Wds. Developer-s wvill nieed to address this over-lap by' adj usting these valuies in a m~anner that eansure-s, a logical escalation in the ECL.The1 "Site spLecific do)Se receptor point" is the distance(s) and/or locations uIsed b5' the licensee to distiniguish betv.'e~n on; site- and offsite doses. The selcte distance(s) aiid'or locations should r-eflect the content of the emferg~ency' plan, and the procedural methodology' used to deter-mine off-site doses and Proetective Actioni Recommendations. The variationl in selected dose recepto ponsmas there may be some differencees in the distance from the release poinit to the calclae-d-doase point from site to site.Developers should researc. radiation monitor design documents Or other inform.....ati.on sources to ensure that 1)the EAL value being considered-d is wit;hin; the usable response and display'i angge of the instrument., and 2) ther-e are no automatic featurfes that may' render the m~onitori reading invalid (e.g., an auto purge feature triggerfed at -a pariciular indicationi level).it is recogniz~ed that the condition described by this WC may resuilt inl a radiological effluent value beyond the operating Or display' rang,=e of the installePd_ efflent moio.In those cases, EAL values should be determined with a margin sufficient to ensure that an accurfate monitor reading is available. For example, an EAL monitor r-eadinig might be set at 909% to 9-5%4 of the highest accurate monitorf re-ading. This provision notw~ithstanding. it 35 1P a iz e theestmatd/clcuate moiteir reading is gr-eater than approximately 1109% of the highest aecur'ate iiion itor reading, then developers may sli dse not to include the monitor as an indieation mid identity an alterate EAL difeshekk. ~Althoulgh the WC refe~rcnes TEDE, tield surfvey resuilts arc generally available einly as a "whole body" dose rate.or-E this reason, the field sur-vey RAL specifies a "closed 'vlindow.," ur'.'Hey reading.Indic;#atios froAm A real time@ dEse projection system are 110t incflude-d-in the generic EAI~s. Many licetnse-es, do not have this capability. For those that do, the eapability may not be within the scope of the plant TFechnicr-al Specifications. A licensee may r~equest to include an EAL using real time dose proejecto ytmrsls approval will be considered an a case by case basis.Indications froem a perimeter menitoring system are not included in the genleric EALs. Many licensees do not have this capability. F-Or those that do, thiese mnonitorFs may not be controlled and- maint~ained to the same level as plant equipment, or withini the scope o~f the plant Technicial Specifiations. In additioni, r-eadinigs miay-4be influenced by enviroenmeantal or other factor-s. A licensee may request to include an E=AL using a per-imeter-moinitoring system;3 approeval will be considered oni a case by case basis.ECL Assignglment At:ri4hit;"A I 1 I -, --36 1 P ag e AA2RA2 ECL: Alei4-ALERT Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel.Operating Mode Applicability: A-gALL Example Emergency Action Levels: (1 or 2 or 3)(1) Uncovery of irradiated fuel in the REFUELING PATHWAY.(2)a. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of the following FHB radiation monitors readings:* FHB Exhaust, RT-8035 or RT-8036 greater than 1.00 E- I LtCi/crn 3 OR* ARM (68' FHB), RE-8090 greater than 1.500 mR/hr OR b. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of the following RCB radiation monitor readings (Mode 5 or 6 only).0 ARMs (68' RCB), RE-8055 or RE-8099 re'eater than 850 mR/hr.NOTE EAL-3 is 17ot applicable u1ntil the enhanced SFP level instrumentation is available for use.(3) Lowering of spent fuel pool level to (site lower.po,,ifie 2 value). [See Dev'lopior 49'- 10" or Basis:-This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool (icc Develpeir Notes) or Inside Containment Storage Area (ICSA).-These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HUI.Escalation of the emergency would be based on either Recognition Category A-RA-R or C ICs.371 P a Re EAL #1- This EAL escalates from AU- RU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations. While an area radiation monitor could detect an in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.EAL #2- This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel.Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident). EAL #3- Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.Escalation of the emergency classificatin le-el EMERGENCY CLASSISFICATION LEVEL would be via ICs AS-1-RS I or A8-RS2 (sc 1S2 A T , -:;'s RA2: EAL-2 Selection Basis: The calculated airborne source term and radiation monitor responses for a fuel handling accident in the FHB is based on Calculation STPNOC013-CALC-005 Rev. 1. The threshold value of 1500 inkhr for area radiation monitor RE-8090 was truncated less than 4% from the calculated value to ensure the threshold was readily assessable. Threshold values for FHB Exhaust Monitors RT-8035 and RT-8036 were also included because they are accident monitors that are sensitive to noble Pases which are expected to be present if irradiated fuel is damaged. The calculated monitor reading for RT-8035 and RT-8036 is 3.8 uiCi/cm' and the high range of the monitors is 0.3 utCi/cm-. The threshold value of 0.1 ,.Ci/cmn is approximately 6 orders of macnitude above background and indicative of damaged irradiated fuel. It was selected because it is readily assessable and within the calibrated range of the monitors.The calculated airborne source term and radiation monitor response for a fuel handling accident in the RCB is based on Calculation STPNOCO13-CALC-005 Rev. 1. The threshold value of 850 mR/hr for RE-8055 and RE-8099 was truncated less than 2% from the calculated value to ensure the threshold is readily assessable. 381 P a g e RA2: EAL-3 Selection Basis: Spent Fuel Pool level of 49'- 10" (Level 2) is a site specific level based on the guidance provided in NEI 12-02.Revision 1, Industry Guidance for Compliance with NRC Order EA-12-051. "To Modify Licensees with Regard to Reliable Spent Fuel Pool Instrumentation", August 2012.In NRC Order EA-1 2-051 and NEI 12-02, Level 2 is defined as the "level that is adequate to provide substantial radiation shielding fbr a person standing on the spent fitel pool operating deck ... " The STP UFSAR identifies the top of the Spent Fuel Storage Racks at 39'- 10". The guidance in NEI 12-02 indicates that 10' of water above the top of the Spent Fuel Storage Racks provides substantial radiation shielding. Ten feet of water above the Spent Fuel Storage Racks is 49'- 10", the threshold value for this EA L.Reference 6 identifies the site specific levels of the proposed SFP level instrument and identifies the Level 2 criteria as 49'- 10".

REFERENCES:

1. Calculation No.: STPNOCO I 3-CALC-005 Rev. 1, Fuel Handling Accident Monitor Response for Emergency Action Levels.2. UFSAR. Rev. 16, Section 9.1.2.1 (SFP Rad levels)3. UFSAR, Rev. 16, Section 9.1.2.2 (SFP top of Racks)4. NRC Order EA-12-051 (SFP levels)5. NEI 12-02. Rev. I (SFP levels)6. South Texas Proiect (STP) Overall Integrated Plan for Implementation of Unit I & Unit 2 Spent Fuel Pool Level Instrumentation to Meet NRC Order EA-12-051, Rev. 0, NOC -AE-13002959 Developer-Notes: FerE~AL #1 Depending upon the availability and range of inistrumientation.

this EAL may include specific readings indicative of4f@e uo eeR;cnSider wAtelad- r-adiatiOnlevl tOE4readings. Speeify, the made applicability 4fa particular indircation; if it ik not available in All m;Ade For EAL #2 The "4site specific listing of r-adiation monitor-s, and the assocei-ated-reaadings. setpoints and/lor a~alarms" shouild contain those radiation monitors that could be used to identif' damage to an irriadiated futel assembly (e.g., ecnfirmFator-y of a release of fssioni pr-oduct ,gases fromi irradiated fuel.).Far EALs # 1 an~d ;V'Dev~elopers should researchi radiation monitor designi documents or other infor-mation sourccs to cnsurce that 1)the E~A.L value being considered is within the usable response and display ranige of the instrument., and 2-) there ae no automatic featur-es that mfa:, render the monitor r-eading invalid (e.g., an auto purg-e featur-e tr-iggered ata particular indication level).39 1 P a a e it iS reW"gnizeOd that !he Econditio deascribed by. this WC may result in a r-adiation Nvalue beyand th@e perating," OF display ranige of the inistalled radiation monitor-. hin those eases, EAL vausshould be deemndwithamri sufflieent to ensurc that an aeeurate mionitor, reading is ava~iable. For examnple, an EAL mnonitor r-eading might be set at 90% to 95-% of the highest accurate moit)Ror readinkg. This provision notwithstanding, if the estimated/calcul-atead moanitor reading is grFeater-than approxtim~ately 1101% of the highest aceurate monlitor, reading. theni developer-s may choose not to include the monpitorf as an indic-ation -and idontitA' an alternate EAL thireshe14. To further promote aeecurate classification, developer-s sheuld consider if same combination of monitors could be classification assessme.it. PRA; MAR;;IL[ +ArI ..[ AppZ, A+1:3 .1, Development of the EALS shIould also consider the availability and limitations of mode dependent, Or other conHtrolled but temipEoraf~'. Fadiation moneiitor-s. Specify the mode applicability of a paricuelar-monitorF if it is not avaPi]Alabl in; All modes.in aeccordance with the discussion in Section 1.4'L NRC Or-der EA 12 05 1. it 29is ecommended that this EAL be implemented whben the enhanced spent fbel pool level instrumwentation is available for- use. The "site specific Level 2 value" is uisually, the spent fuiel peal level tha! is adequate to provide substantial radiation shielding for a person standing on the spent fuiel pool operating deck. This site specific level is determined in aecofdanc-e with NRC Order EA 12 05 1 anid NEI 12 02. and applieable own...r' s group guidance.D~evelopers Should modiA, the EAL and/or- Basis section to reflect any, site specific conistr-aints or lim:itationis assciaed iththedesgn r per.ationl of instrumnenltat ion used to determine -t -Lee 2 vau.-Eel= Assignment Atr~ibutes: 3.1.2.B and 3.1.2.G 40 1 P a R e AA3RA3 ECL: Alei4 ALERT Initiating Condition: Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.Operating Mode Applicability: A-IALL-E-xampie-Emergency Action Levels: (1 or2)Note: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. (1) Dose rate greater than 15 mR/hr in ANY of the following areas: " Control Room ARM (RE-8066)OR" Central Alarm Station (CAS) by radiation survey" (ether site specifie arcas,'rocms) (2) An UNPLANNED event results in radiation levels that prohibit or impede access to iyANY of the following plant racms. or areas listed in Table H3/R2: TABLE H3/R2: Plant Areas Requiring Access RCB RHR Heat Exchanger Rooms o *1 MAB 51 ft Room 335 EAR Roof, MCC 1G8, 4.16KV Switchgear Rooms o "nl EAB 4.16KV Switchgear Rooms 41 1Pa e (site specifi list of planit rEms or areas with entry related mode applieability identified) Turbine Genierator BEuildin (TGB)lselation Valve-Cbee(V Fuel aninBilig(Fl-B) Reacter Containment Building (CB Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the radiation levels and determine if another IC may be applicable. For EAL #2, an Aler-tALERT declaration is warranted if entry into thý affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the iner-eased higher radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).An emergency declaration is not warranted if any of the following conditions apply.* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode I when the radiation i-nefeaserise occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.* The infeeasedhbgher radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).* The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections)." The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.Escalation of the emergency classification level EMERGENCY CLASSIFICATION LEVEL would be via Recognition Category_-R, C or F ICs.RA3: EAL-1, EAL-2 Selection Basis: The NEI 99-01 value of 15 mR/hr is derived firom the GDC 19 value of 5 remn in 30 days with adiustment for expected occupancy times. The rooms listed in EAL-1 require continuous occupancy to maintain normal plant oneration. or to nerform a normal cooldown or shutdown.421 P a e The areas listed in EAL-2 apply to areas that contain equipment necessary for plant operations, cooldown, or shutdown. Assuming all plant equipment is operating as designed. Normal operations and safe shutdown equipment operation is capable from the Main Control Room (MCR). Tile plant is able to transition into a hot shutdown firom the MCR, therefore H3/R2 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary following entry into hot shutdown (establish Residual Heat Removal shutdown cooling. disable operation of charging and ECCS equipment, and limit dilution pathways) and subsequent entry into cold shutdown (disable operation of ECCS equipment). After achieving cold shutdown it is assumed that the plant will be maintained in a cold shutdown condition.

REFERENCES:

1. General Design Criteria 19 2. OPOP03-ZG-0008, Rev. 56. Power Operations

3. OPOPO3-ZG-0006.

Rev. 54. Plant Shutdown from 100% to Hot Standby 4. OPOP03-ZG-0007. Rev. 71, Plant Cooldown DeveepF Netesi EAL 4!The value of lm'hr is deriv~ed from the GDC- 19 value of 5 rem in 30 days v.'ih adj uStffenlt for eXpected eceupan y tms The ....ther Site speeific arieas.rooms" should in..lude any areas or rooms requiring contin-ous occupancy to mIaint-ain normal plant operation, or- to perfor;m A nor-mal -an ShutWdnA". The "site specific list of plant rooems or ar-eas with enitry r-elated mode applieability idenitified" should specify those rooms or- areas; that contain equipment whichl require a manual/loc-al action as spccified in operating procedures used for- normal planit oper'ationi, eooklown and shutdown. Do,niot include roams Or areas ill whic-h actions of a contingent Or emerg-ency natur-e wouild be perfor-med. (e.g., an action to address an cff normal or emergency, eonditiosi suchi as emergency repairs, correctiv~e measurfes Or emer-gency operations). in addition, the list should spec-ify the plant mode&(s) during whichi entr-y wouild be requir-ed for eachi r-oom or ar-ea.The list Should not iAelude ro.oms or areas for. which entry, is required solely to per-form actions of an administrative or recod keeping nature (e.g.. neal romds Futin-e inspections). if the equipmienit in the listed room or ar-eawas already inoeper-able, or etut of service, befor-e the evenlt occuirred, then no emergency should be declar-ed sincee the event will have tio advere impacat beyond that alr-eady allowed by Technic-al Speeif~cations at# the time of thea event.& T = I Nooems -andi -Areas listedi in 4--. 6 ;. i A30 noAt needi to be mnemusmeg in FL/\L 172. mnieudimng Iime GontrolI Koo.-Aa ECL Assignment Attributes: 3.1.2.C NEI 99 01 (.Revision

6) Novemiber 2012 11 43 P a R e ASI-RS1 ECL: Site Area Emergency SITESITE AREA EMERGENCY Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 100 rnrem TEDE or 500 mremthyroid THYROID CDE.Operating Mode Applicability: A4UALL Exam pie Emergency Action Levels: (I or 2 or 3)Notes:-The Emergency Director should declare the Site Area EmergencySITE AREA EMERGENCY promptly upon determining that the applicable time has been exceeded, or" will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.e If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer-vat4 VALID for classification purposes.* The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. (1) Reading on imyANY of the following radiation monitors greater than the- eadin values listed in Table RI column "SAE" shown for 15 minutes or longer: (site speeifie monitor- list and thrfeshold valueso Table RI: Effluent Monitors Release Point Monitor GE SAE ALERT UE Unit Vent RT-8010B 1.50 E+08 iiCi/sec 1.50 E+07 ttCi/sec 1.50 E+06 kiCi/sec 1.40 E+05 tiCi/sec Main Steam RT-8046 tirt 4.00 E+02 itCi/cm 3 4.00 E+01 yCi/cm 3 4.00 E+00 ktCi/cm 3 5.00 E-02 [tCi/cmn 3 Lines 8049__.___ (2) Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyfeidTHYROID CDE at or beyond (site specific dose receptor point) the SITE BOUNDARY.(3) Field survey results indicate EITHER of the following at or beyond (site specific dose receptor point)the SITE BOUNDARY: " Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate hfeid-THYROID CDE greater than 500 mrem for one hour of inhalation. 44 1 P a P e Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA ProtecetiveAetion Guides PROTECTIVE ACTION GUIDES (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem4hyreid THYROID CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and-t4wreid THYROID CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid -VALIDVALID for classification purposes.Escalation of the1.emerzenc;v celassiatiton level EMERGENCY CLASSIFICATION LEVEL would be via IC RSI: EAL-1 Selection Basis: The Unit Vent and Main Steam Line monitor readings were included in this EAL because they give instantaneous indications of a monitored gaseous release meeting the EAL threshold values of 100 mrem TEDE or 500 mrem CDE THYROID at the SITE BOUNDARY. Gaseous releases from the plant are monitored by the Unit Vent. The Main Steam Line monitor readings correspond to a concentration that would result in a release rate meeting the EAL threshold values if there were a release via the Power Operated Relief Valves (PORVs) or Safety Relief Valves.The Unit Vent and Main Steam Line release values are based on Calculation No. STPNOCO13-CALC-002 Rev.2. The adiusted values used in this EAL were conservatively truncated by less than 1% of the calculated values to ensure they are readily assessable. RSI: EAL-2. EAL-3 Selection Basis: N/A

REFERENCES:

1. Calculation No: STPNOC013-CALC-002 Rev.2, Radiological Release Thresholds for Emergencv Action Levels 2. UFSAR Section. Rev. 14. Section 11.5.2.3.3 and 11.5.2.5.3 (monitor descriptions)

Dcvclopcr)e Notes: While this IC may not be met absent ehallcnges to multiple Assion proeduct barrier-s, it provides elassirieatiin diversity and May be used to elassifv events that would noAt rc-ach the Same 9(CL based en plant ztatuS Or the-isc produc-t matrixi alonle. FEr mfany of the D8As anialyzed in the Updated Final Safeaty Analysis Report, the 451 P a g e discime-oin;ator; will not bpe the. numbier of fission producet barr-iers challeanged, but r-ather- thea Amount of radioactivity Frele,-kaed-to- the evrnet The EPA PAGs are expressed in terms of the sumi of the effeetive dose equivalent (EDE) and the committed effctie dseequivalenit (CEDE), or as the thyroid committed dose equivalent (C-DE). For the purpose of thes IC/EAkks. the dose quantity total effective dseuvAln (T-EDE), as defined in 10 CPR § 20, is usead in lieu of Ae.ii f EDE andCE .The EPA PAG guidance proevides for the uise of adult thyr-oid dose conversion factors; however, some states hav.e dcddto b-ase proatective ac-tions oni child thyr'oid C-DE. NJucleaF power plant !C~s/Elg~s need to) be conlsistent with the proetective action methodologies em ple.,ed by the States within their- EPZs. The thyrFoid C-DE dose used in the IC and EAL~s should be adjusted as necessar-y to align with State protective action decision maki-ng The "site specific monitor list and thr~eshold values" should be determiined with consideration of the followingi" eeietten of-tnc appro)priate inStalfeo gaseous ef-Buent fmonitors." T-he effluent moitorE readinlgs Should correspond to a dose of 1 00 mrcm TEDE or 500 mr-em thyro~id CDE; at the "site specific dose receptor-point" (consistent with the calculation methodology, employed)forF one hourf Of eXposure._ MoA-nitorf r.eadings will be c-a cauated asing a set o .assum.ed metee-eological data o" atmoheric dispersion fac-tors; the data or factor-s selected forf use should be the sam~e as those employed to calcuilate the monitor- readinggs for !Gs AA 1 and AG 1. Akcceptable sou-ces of+this information include, but are not limited to, the RETS/ODC-N and values used in the site's emergency dose assessment methodology." The calculation of monitor readings will also r-equir-e use of an assumed release isotopic mi*x; the selected mix should be the same as that employed to calculate mon1f.itor. r..adi..gs fr- WTs AA 1 and AGA.A 1 sources of this hIIn matLI I bLit are not limited E a, the RETS/ODCM a'nd values used in thp. 4pt'A; sourpres'; d.ow;.p .s ,. msthdlov.-1 Depending upon the jnethodology, used to calculate the EAL values, there may be overlap of some valuies between differ-ent W~s. Developers will need to addr-ess this overlap by adjusting these values in a manner- that ensur~es a logical escalation in the ECL.The "site speehifi e---- receRptor point" is the distance(s) andor locations used by the lieenseeto distinguisTh between on site an;d- offsite doses. The selected distancee(s) an~bJor locationwSs shoulmId-reflect the conitenit of the emergenoy, plan, anld the proce~dural methodology used to determinife offSitea doescu And Proatective Acto Recommendations. The variation in selected dose recepto ponsmas there may, be som~e differences in the dist-ance froam the release point to the c Alculated dose point fr-omf site tosie Developers should research radiation monitor- design documents Or oth1er information sources to ensure that I)the- E-AL- v-alue being conlsidered is W.ithin the uisable response anid display' range@ ofhe instruEment, and 2) there are no auitomatic featur-es that mnay r-ender- the moenitor-readinig invalid (e.g., an auto pafrge feature trigger-ed at a par-ticularindcto level).it is recgnze tht h cnition descPribed by, this WC may, r-esult in a radiologial eff~luent v'llalu beyond the operating or display range o~f the installed eft44fent moitor. fInthse c-ases. EAL values shouild be determ~ined with a margi suficen to ensure th.t an accuate monitor- reading is av~ailab-lea. For example. an ALI monitor÷1eadin miught. be set at- t'pljaýLo ,.1; ..-... A-f ;T .... R igeSt : .. i ;L... ÷ r 1uuH1O. I H P-I ,tA.i....ajilIf,",, i the estimated/calclated monitor reading is greater than apprOXimatel' 11090; of the highest accurate monfitor 46 1 P a ge redig then devlapr ma hos o t nluete .oio as an. indicatiion mnd identify, an alter-nate EAL threshold~Although the lC reforenees TEDE, field survey resuilts are generally available only as a "whole body" dose Fate.For this reasoni, the field sur-vey EAL speeifies a "e. I sed winidow" survey reading.inidications from a r-eal time dose proejection systeim are not ineluided in the generic EAbs. Many licensees do not. ave hIs -capability. F-Or those that do, the eapabilit,' may, not be within the scope of the plant Technical Specifications. A licensee may r~equest to include an EAL using real timne dose pr-ojection system results;appr-oval will be conisidered en a case by case basis.hi~dicat ions from a per-imetcrmntrigsse are fi Ot included in the generic EALs. Many lieen: sees dE- Ret ha'.'ei may HOt fie CONIFOlled and maintailied te she s-ame level as planlt eqUipmient, orf within the scope of the plant Technical Specifications. in additiotn, readings miay be influenced by eirn entaloother factors. A licensee may r-equest to include an EAL using a perimeter-monitoring system; approval will be consider-ed on a case by case basis.ECL Assiginmenit Attributes: 3.1 .3 .C 471 P a e AS9RS2 ECL: Site A:"ea Emergency SITE AREA EMERGENCY Initiating Condition: Spent fuel pool level at (site .peeifie Level 3 de.c.-iptin) lower.Operating Mode Applicability: AI4ALL E-taniple Emergency Action Levels: NOTE EAL-1 is not applicable until the enhanced SFP level instrumentation is available fbr use.(1) Lowering of spent fuel pool level to (site .pe.ifie Level 3 desceription)40'-4" or lower.Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency SITESITE AREA EMERGENCY declaration. It is recognized that this IC would likely not be met until well after another Site Area Emergency SITESITE AREA EMERGENCY IC was met; however, it is included to provide classification diversity. Escalation of the emergency elassifieatien leve- EMERGENCY CLASSIFICATION LEVEL would be via IC AG-G-RG I or AG-. RG2.RS2: EAL-1 Selection Basis: Spent Fuel Pool level of 40'- 4" (Level 3) is a site specific level based on the guidance provided in NEI 12-02.Revision 1. Industry Guidance for Compliance with NRC Order EA-12-05 1. "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", August 2012.In NRC Order EA-12-051 and NEI 12-02. Level 3 is defined as "level where fuel remains covered and actions to implement make-up water addition should no longer be deferred. " The STP UFSAR identifies the top of the Spent Fuel Storage Racks at 39'- 10".Reference 4 identifies the site specific levels for the proposed SFP level instrumentation and identifies the Level 3 criteria as 40'- 4".481 P age

REFERENCES:

1. UFSAR. Rev. 16, Section 9.1.2.2 (SFP top of Racks)2. NRC Order EA-12-051 (SFP Levels)3. NEI 12-02, Revision 1, Industry Guidance for Compliance with NRC Order EA-12-051. "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", August 2012 4. South Texas Proiect (STP) Overall Integrated Plan for Implementation of Unit I & Unit 2 Spent Fuel Pool Level Instrumentation to Meet NRC Order EA-12-051, Rev. 0, NOC -AE-13002959 Developer Notes: hi aeear-dance with the discussion ini Section 1.1. NRC Order- EA 12 05 1, it is r-ecommended that this WC and EAL be implemented when the- enh11anceRd-Spent fuel pool level-0 ins.,-t.rumen;4.--;;ta;-tion-4 is lavailaible.

for. use. The '"site specific Leel 33 valu'il" is us.ll:,' that spent fu.el pool level where fuel r ,emains ..vered and aetions to implement make up water addition should no longer be deferred. This site specific level is determiined in accordance with NRC Order EA 12 051 and NEI 12 02, and applieable -owner's g.r--p guidance.DeeoN'1pers shOUld mo1dify the- EAL a1d- o Ba-siSetn to) reflet assciated with the design or operation of instrumentation used to aR site speeicife on4str-air

1. tS or limitations kiP+pR!!fl:lp-

+QP f. , -".,Pi .VAI-P-rCT A ch: -I x,. I 21 491 P a ge AG4RG1 ECL: General EmergeneyGENERAL EMERGENCY Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrent-hyre44 THYROID CDE.Operating Mode Applicability: AIIALL lAi-Fmple-Emergency Action Levels: (I or 2 or 3)Notes:* The Emergency Director should declare the General Emergeency GENERAL EMERGENCY promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer val4i-VALID for classification purposes.* The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. (1) Reading on aiiyANY of the following radiation monitors greater than the-.eadin values listed in Table RI column "GE" reading .h.on, .for 15 minutes or longer: (sthe speeifie moenitor lit and thresholdv;alues) Table RI: Effluent Monitors Release Point Monitor GE SAE ALERT UIE Unit Vent RT-8010B 1.50 E+08 t.Ci/sec 1.50 E+07 itCi/sec 1.50 E+06 LQCi/Sec 1.40 E+05 utCi/sec Main Steam RT-8046 thru 4.00 E+02 LOi/cmn 3 4.00 E+01 tCi/cm 3 4.00 E+00 ICi/crn 3 5.00 E-02 itCi/crn 3 Lines 8049 (2) Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem THYROID CDE at or beyond the SITE BOUNDARY.(3) Field survey results indicate EITHER of the following at or beyond (site specific dose r.ecepter Point)the SITE BOUNDARY:* Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.,OR* Analyses of field survey samples indicate hyt-ei4 THYROID CDE greater than 5,000 mrem for one hour of inhalation. 501 P a e Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Pretective Action Guides PROTECTIVE ACTION GUIDES (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 rnrem-hyf-id THYROID CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE andthy-r-id THYROID CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer-a4i4 VALID for classification purposes.RGI: EAL-1 Selection Basis: The Unit Vent and Main Steam Line monitor readings were included in this EAL because they give instantaneous indications of a monitored gaseous release meeting the EAL threshold values of 1000 mrem TEDE or 5000 mrnrm CDE THYROID at the SITE BOUNDARY. Gaseous releases from the plant are monitored by the Unit Vent. The Main Steam Line monitor readings correspond to a concentration that would result in a release rate meeting the EAL threshold values if the release was via the Power Operated Relief Valves (PORVs) or Safety Relief Valves.The Unit Vent and Main Steam Line release values are based on Calculation No. STPNOCOI3-CALC-002 Rev.2. The adjusted values used in this EAL were conservatively truncated by less than 1% of the calculated values to ensure they are readily assessable. RGI: EAL-2, EAL-3 Selection Basis: N/A

REFERENCES:

I. Calculation No: STPNOCO 1 3-CALC-002 Rev.2, Radiological Release Thresholds for Emergency Action Levels.2. STP UFSAR, Rev. 14, Section 11.5.2.3.3 and 11.5.2.5.3 (monitor descriptions) [Sec D.v..opeF Notes]51 1Page AG2RG2 ECL: General Emergeci'yGENERAL EMERGENCY Initiating Condition: Spent fuel pool level cannot be restored to at least (site specific Level 3 deseripticn)40'- 4" -for 60 minutes or longer.Operating Mode Applicability: AIIALL Example Emergency Action Levels: Note: The Emergency Director should declare the General En'eigeneyGENERAL EMERGENCY promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.NOTE EAL-1 is not applicable until the enhanced SFP level instrumentation is available for use.(1) Spent fuel pool level cannot be restored to at least (site specific Level 3 des_. iptionL.. -4" for 60 minutes or longer.Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment. It is recognized that this IC would likely not be met until well after another General Emergnc:'y GENERAL EMERGENCY IC was met; however, it is included to provide classification diversity. RG2: EAL-1 Selection Basis: The Spent Fuel Pool level of 40'- 4" (Level 3) is a site specific level based on the guidance provided in NEI 12-02, Revision 1, Industry Guidance for Compliance with NRC Order EA-1 2-051. "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", August 2012.In NRC Order EA-1 2-051 and NEI 12-02. Level 3 is defined as "level where fuel remains covered and actions to implement make-up water addition should no longer be deferred. " The STP UFSAR identifies the top of the Spent Fuel Pool Racks at 39'- 10".Reference 4 identifies the site specific levels of the proposed level instrumentation and identifies the Level 3 criteria as 40'- 4".52 1P a Re

REFERENCES:

1. UFSAR, Rev. 16, Section 9.1.2.2 (SFP top of Racks)2. NRC Order EA-12-051 (SFP Levels)3. NEI 12-02, Rev. 1. Industry Guidance for Compliance

\vith NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", August 2012 4. South Texas Project (STP) Overall Integrated Plan for Implementation of Unit I & Unit 2 Spent Fuel Pool Level Instrumentation to Meet NRC Order EA-12-051, Rev. 0, NOC.-AE-13002959 Dev.eloperiNotes. in aeccrdance with the discussion in SeItion 1.4, NRC Order EA 12 05 1, it is recommended that this Ic and EAL be implemented when the enhanced spent fuel pool level instrumentation is available for uise. The "site spec-ific Level 3 value" is usually that spent fuiel peel level wher-e fuel r-emainis covered -and atnsto implemenit mfake Lip water addition should no longer- be deferred. This site specific level is deter~mfined in accor-dance with O.4Kk "uraer Iz,. 4-6 UZ) an EIHE i..f~ !UIN and appiteable owner:s group gudagnee.Developers should moadify, the EAL andir Basis section to reflect any site specific constr-aints Or limiitations associated with the design or oper-ation of instrumentation used to determnine the Level 3 value.ECL Assignment Attrhi~~~1 1 4lfG 53 1Page 7 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS Table C-i: Recognition Category "C" Initiating Condition Matrix UNUSUAL EVENT CUI UNPLANNED loss of (Feaet-v~esse1,'RCS [PUW] i-e R4 [B;MR) r inventory for 15 minutes or longer.Op. Modes: 5.6C-6/4 Shitkldwii, 5n6 ALERT SITE AREA EMERGENCY GENERAL EMERGENCY CAI Loss of --eaete+vessel RCS [4WR-] er-inventory. Op. Modes: 5.6 Ce4/d CS1 Loss of (+feaet-vessel RCS -rJ% te.RPV -.B4R4.) inventory affecting core decay heat removal capability. Op. Modes: 5_6C-ld Shutdoqwi, Refuoelhng 5,6 CG1 Loss of (eaetef vessel RCS [Pff;R] or RPV[-WR]) inventory affecting fuel clad integrity with containment challenged. Op. Modes: 5,6C-m Shtidow;'n, Refiuolig 5,6 CU2 Loss of aI4ALL but one AC power source to emergency buses for 15 minutes or longer.Op. Modes: .56,6CdG Shfikdown, Refuoling"-5-Defueled CU3 UNPLANNED ineefeaserise in RCS temperature. Op. Modes: 5.6C-e-d Shutdown, Rei4ding 5,ý6 CU4 Loss of Vital DC power for 15 minutes or longer.Op. Modes: 5.6C-oek S/htdo'wn, Ro.,,ing 5,6 CA2 Loss of a#ALL offsite and aI4ALL onsite AC power to emergency buses for 15 minutes or longer.Op. Modes: 5,6 C-64d Shuitdown, R.f. ,;ing-5,, Defuteled CA3 Inability to maintain the plant in cold shutdown.Op. M'odes: 5.6C-44RDf,.,lIg 5ý6 CU5 Loss of allALL onsite or offsite communications capabilities. Op. Modes: 5,6. G-4 45--Defueled CA6 Hazardous event affecting a SAFETY SYSTEM needed for the current operating 541 P a e mode.Op. Modes. 5_6 C-44 Shz~t~deivi, Refiieing 5,6 55 1 P a g e cul MLA/ oti:,icaiI'cn C f Unusual Event UNUSUAL EVENT Initiating Condition: UNPLANNED loss of(reaeter "esse!/RCS 'DPWR] , r RPV [BWR)] inventory for 15 minutes or longer.Operating Mode Applicability: Cold Shutdown, Refucling 5,6 Example Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the Unusual EventUNUSUAL EVENT promptly upon detennining that 15 minutes has been exceeded, or will likely be exceeded.(1) UNPLANNED loss of reactor coolant results in (i.e.e.e..-v.ese..RCS[PMr] or RPM [B4R]) level keee than a r.equired lower lim.it below the Reactor Vessel Flange procedurally required limit for 15 minutes or longer.(2) a. Reaeter 'el/CS [PWR] or RP rWkJ.]) level cannot be monitored. AND b. UNPLANNED inefeaserise in (SITE SPECIFIC SUMP AND/OR TANK) ANYANY of the following sump or tank levels in Table C2:, Table C2: RCS Leakae" Containment Normal Sunmp* Pressurizer Relief Tank (PRT)* Reactor Coolant Drain Tank (R.CDT)" MAB Sumps 1 thru 4* Containment Penetration Area Sump" SIS/CSS Pump Compartment Sump Basis: This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor (reaeter -essel/RCS [PJJ'R] or RPV [8WIR]) level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.Refueling evolutions that deereaselower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event UNUSUAL EVENT due to the reduced water inventory that is available to keep the core covered.56 1P a g e EAL #1- recognizes that the minimumn required ([reaet. ,i vessW rC " o["! @!- RPV [BURj) level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the mininmum1 level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typieally typieally specified in the applicable STP operating procedure but may' be specified in another. controllin. g docuiment. The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.EAL #2- addresses a condition where all means to determine (Feaetoer. vesse.1RCS r 4.j O] r RPV [B4JrI) level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reaetor vessel/RCS rP-WR] OF -fPV Continued loss of RCS inventory may result in escalation to the Ale'4-ALERT emergenicy classification le-el EMERGENCY CLASSIFICATION LEVEL via either IC CAI or CA3.CUI -EAL-1 Selection Basis: Normal reactorv-essel!RCS inventory duri:ng refueli-. otazes is maintained above the reactor vessel flange during refueling outages Ptper OPOP03-ZG-0007, Plant Cooldown.. RCS level below the ;essel flange is to be minimized. RCS level may be dr-opnjglowered below the vessel flange for specific purposes *-rdei controlled evolutions (e.g.. head removal, mid-loop operations) as described in OPOPO3-ZG-0009, Mid-Loop Operation. The 15 minute time frame allows for prompt operator actions to restore RCS level in the event of an UNPLANNED lowering of RCS level below the operating limit.-CU1 -EAL-2 Selection Basis: This EAL includes two conditions. The first condition is the inability to monitor RCS level and the second condition provides secondary indications that inventory loss may be occurring. The secondary indicators of inventory loss include a list of tanks/sumps found in OPOP04-RC-0003. Excessive RCS Leakage. Since other system leaks could rise levels in various tanks and sumnpl)S. the list has been limited to the tanks and sumps that would have the highest probability of indicating RCS leakage inside the Reactor Containment Building.Although procedure OPOP04-RC-0003 is desimnated for use in modes 1-4, its logic is applicable to this EAL.

REFERENCES:

1. OPOP04-RC-0003, Rev. 18. Excessive RCS Leakage 2. OPOP03-ZG-0007, Rev. 71, Plant Cooldown 3. OPOP03-ZG-0009, Rev. 59. Mid-Loop Operation Devlper- Notes!EAL -1 It is recognized that the minim.um allowable r1eactor..-.esse/R.S/.RPV level may have many values over. the course of a refueling outage. Developers should solicit input f..o. licensed perators concerning the optimum wor-ding for this EAL statement.

hi particular-, determine if the geei oding is adequate to ensur~e 57 1 P a g e acuawie and timely classificatiten. E if speoif; setpoints. .an hbe. inc.uded without m.aking. the EAL stateme1n unw.ieldy-Or p,. ent.ially insie. Si.ten.t with a.tion.s that may be taken, dring..... an otage. if specific setpOin.. s a. e included, these should be dr-awn fromff applic-able operating, proedwre" et-othei-eentfolhing dacumens EAL 42.b Enter any "site specific sump and/or tank" levels that c:uld be expected to inrease ifthere werea loss of invento:ry (i.e., the lost inventory w"ould e R t L. F thle listed Sump a!. otank).A "I .1 1 A 1-44 ai~ssieimrnent Altrietites: 1.1..581Page CU2 ECL: Nctificatien of Unusual Event UNUSUAL EVENT Initiating Condition: Loss of atALL but one AC power source to emergency buses for 15 minutes or longer.Operating Mode Applicability: Gold Sl. td.wn, Reft;elinfa.. 6, Defueled&xan+jle Emergency Action Levels: Note: The Emergency Director should declare the Unusual Evefnt UNUSUAL EVENT promptly upon detennining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. AC power capability to (site .p..ifi. -m-i-geney busce) "IALL three 4160V AC ESF Buses is reduced to a single power source for 15 minutes or longer.AND b. ANY additional single power source failure will result in loss of OIALL AC power to SAFETY SYSTEMS.Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of allAC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Ak-4 ALERT because of the incFteasedadditional time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. genm-e examplee-Examples of this condition are presented below." A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator). e A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being baek-fed from the unit main generator." A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being bae-k-fed from an onsite or offsite power source.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.The subsequent loss of the remaining single power source would escalate the event to an Alert ALERT in accordance with IC CA2.59 1 P a P e CU2: EAL-1 Selection Criteria: The condition indicated by this EAL is the degradation of the offsite and onsite power systems such that any additional single failure would results in a loss of all AC power. This condition is an UNUSUAL EVENT during modes 5. 6 and Defueled because of the additional time available to restore power due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. In modes 1-4, this condition is an ALERT as described in SA 1.

REFERENCES:

1. OPOP04-AE-0001, Rev. 44. First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus 2. OPOP04-AE-0004, Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus 3. OPSP03-EA-0002, Rev. 32, ESF Power Availability

4. Drawing OOOOOEOAAAA.

Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. I & 2 60 1 P age EDeN'e!E)e EAL-NeSi For a pewei* sourec that has Multiple gener-ators, the EAL and/or- Basis Sectioni shOUld reflect the m~inimumi~ number of operatini generatE)rS nleceSSary for that sourcee to pro~vide r-equired powNNer to an AC emergency bus.Far- exiample, if a baeckup power sourcfe is compr-ised of two generatOrs (i.e., twoe 50% capacity generatorFS sized toA fiee-d I A.C emer-gency bus), the EAL and Basis section must speei4ý that both generatorS fcr that source ar~e eeiang7 The "site specific emergency5 buses" are the buises fed by off-site or- emergcney ~AG power sourcees that supply powAer to the electric-al idistribuition system that powers SAFETY SYSTEMS. Theare, is typically 1 emnergency bus per: train of SAFETY SYSTEMS.Developers should modify, the buileted examples provided in the beasis section. above, as nede t -fleet their site specific pln des Igns and capabilities.0 The EAbs and Basis shou~ld reflect that eachi indeaenident offste flower circuit eonstitutes a singrle nower: snuree.ro eamle tnrlee mcij~ePenaenlt

jqoj-N o~fsIte powLFer circuts ki.0,le. incoing pow;Aer lines) comprise tnree separ'ate powAer sour~es. lIndepenidence mfay' be determiiined fromi a i-eview of the site specific U2SA.R. SBO analysis or, related less of electrical power: stuldies.The EAL and/or- Basis section may specify tise of a nion safety related power souirce provided that oper'ation et this source is recognized in AOPs and EOPS.ý or beyond design basis accident response aguidelines (e.g-., FLEX Stippei4 guidelinies).

Suceh power: sources should generally m~eet the "Alterniate acG sourcFe" definition provided in 10 C FR 50.2.~At multi unit stations, the EALs may, cr-edit opnstr measures that ar-e procedur-alized an;;d- can be implemented within 15 minut-- es.Cosider capabilitikees suhas power source croess ties, "swing" genierators, othe power sources described in -abnonm.al or emergency operating procedures, etc. Plants that hav~e a proceduralized capability to supply off-site AC pov.'er to an aff-ected unit via a cross tie to) a comI~panionuit acei this power source ini tile EAL provided that the plannled cross tie strategy meets the requrcrncnts of 10 CFR 50.63.IZULý+- ASSIRRnment ,'uRiijut:4i 3.I. I.A 61 1Page CU3 ECL: Notification of Unusual Event UNUSUAL EVENT Initiating Condition: UNPLANNED inefeaserise in RCS temperature. Operating Mode Applicability: Cold Shutdown, Refueling

5. 6 Example Emergency Action Levels: (I or 2)Note: The Emergency Director should declare the Unusual Event UNUSUAL EVENT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) UNPLANNED i-neFeaserise in RCS temperature to greater than (site specific Technical Specification cod szhutdown temperatur.

e limit) 200 'F (Tavg).(2) Loss of ALL RCS temperature and (Feaete: HvesselRCS [.PR] or RPM [BWR]) level indication for 15 minutes or longer.Basis: This IC addresses an UNPLANNED in+e-easerise in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to deterniine RCS temperature and level, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA3.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. EAL #1- involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange.Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid i-*eFeaserise in reactor coolant temperature depending on the time after shutdown.EAL #2- reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation to Alei4-ALERT would be via IC CAl based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.621Page CU3: EAL-1 Selection Basis: An UNPLANNED temperature rise above 200 'F would result in an UNPLANNED mode change due to the inability to control RCS temperature. Mode 4 (Hot Shutdown) would be entered when Tavg exceeds 200 OF (Reference 1).CU3: EAL-2 Selection Basis: N/A

REFERENCES:

1. Technical Specifications Table 1.2 (Mode. Temperature, Power. keff Table)Dels-lopcr Notcs:.... II ..... .... ..... .. ... ..FEor EAL 41., enter- the "site speeti~e Teellhnieal SpeC-ltlcatilon CcAld- ShutdEWn tem:peraiture limlit" Where inidieated.

LLL A5Sl~nrnent Attrltutez: 1.1./'.631 PaPe CU4 ECL: Netimfcatien of Unusual Event UNUSUAL EVENT Initiating Condition: Loss of Vital DC power for 15 minutes or longer.Operating Mode Applicability: Cold Shutdown. Refieling

5. 6 Exa*imple.Emergency Action Levels: Note: The Emergency Director should declare the Unustal E-vent UNUSUAL EVENT promptly upon detenrmining that 15 minutes has been exceeded, or will likely be exceeded.(1) Indicated voltage is less than (site specific bus voltage value) 105.5 VDC on required Vital DC buses for 15 minutes or longer.Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions inereaseextend the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.As used in this EAL, "required" means the Vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment.

For example, if Train A o+and C is-are out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event UNUSUAL EVENT. A loss of Vital DC power to Train A and/or C would not warrant an emergency classification. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Depending upon the event, escalation of the emergency classification level EMERGENCY CLASSIFICATION LEVEL would be via IC CA I or CA3, or an IC in Recognition Category AR.CU4 -EAL-1 Selection Basis: The minimum voltage for Class I E 125 VDC battery buses was determined in calculation 13-DJ-006, Rev. 3 to be 105.5 volts. At 105.5 volts or less, OPOP05-EO-ECOO, Loss of All AC Power, directs the operators to open the battery output breakers.

REFERENCES:

1. Calculation I 3-DJ-006, Rev. 0. 125 VDC Battery Four Hour Coping Analysis 2. OPOP05-EO-ECOO, Rev. 23. Loss of All AC Power 64 1 P ag e Developer-Notcs: The "site spe;ifie bas oEtlytage value" should be based on the ..inimum b. svoltage neessary f.r adequtate Operation of SAFETY SYSTEM equipmfent.

This voltage v~aluie s houLld incOr-porate a margin of at least 1 5 minutes of operationi befor~e the oniset of iinabilit' to opefate these leads. This voltage is usually ncar h minimuim voltage selected when battery sizing is perfrqm~ed-. The typical value fori An etaRire bailer',' set is approximately 105 NIDC. For A 60 cell string of bafteries. the eell voltagge is approximately 1.75 Volts per cell. For a 58 string battery set, the minimum voltage is approxEimately 1.8-1 Volts per cell!.ECL Assignment Attributes: 3.1..1At 65 1 P a g e CU5 ECL: Notification of Unusual Event UNUSUAL EVENT Initiating Condition: Loss ofa44ALL onsite or offsite communications capabilities. Operating Mode Applicability: C-ld Refueliiný-5, Defileled Fxaniple Emergency Action Levels: (I or 2 or 3)(1) Loss of ALL of the following Onsite communication methods in Table C4.(site spe)ific list of e lin tescommunications methds)(4l)L2) LIoss of ALL of the following Offsite Response Oryanization (ORO) commn r ilication methods in Table C4.(site seceitie list ot communications mtheos)(3) Loss of ALL of the following NRC communication methods in Table C4.ksite Sneifie list ot communications maetheds)T2hk C~1 CnmmIInic2tinn~ M~tbnd~EAL-1 EAL-2 EAL-3 ONSITE ORO NRC Plant PA system X Plant Radios X Plant telephone system X X X Satellite phones X X Direct line from Control Rooms to Bay City X X Microwave Lines to Houston X X Security radio to Matagorda County X Dedicated Ring-down lines X ENS line X Basis: This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).EAL #1-addresses a total loss of the communications methods used in support of routine plant operations. 66 P1Paiee EAL #2-addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are (se De;'lepe: Notes) Matagorda County Sheriff's Office, and Texas Department of Public Safety Disaster District in Pierce.EAL #3-addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. CU5: EAL-1, EAL-2, and EAL-3 Selection Basis: Lines not included for offsite communications to ORO and NRC included links that would need relaying of information. Links were obtained from procedures OPGP05-ZV-001 1, Emergency Communications.

REFERENCES:

1. OPGP05-ZV-00 11, Rev. 8, Emergency Communications Deve'loper-Notes: 1F=A -.41 The "site speeific list of com,,iatio methods" should include all ;ommunications meth.As used for rutine planlt commiunications (e.g., mercial or site telephones, page pa.. y syste..:s, radios. etc.). This listing should icueinsctalled plant equipment and components, and not itmAwned an;d maintaine -by EAL #2 The "site specific list of conmmunications methods" should include all communications methods used to per-form inlitial emiergenc-y notifications to OROS as described in the site 59 Emnergency Plan. The listing, shouild include installed plant equipment and compoeniits, and not items owned and maintained by individuals.

Example methods are ring down/dedicated telephone ines, commercial telephone lines, rad ie, satl lite telephones and in temnet based commfunications technology. in the Basis section, inseA the site specific listing of the OR~s r-equir-ing notification o~f an emergency declaratio fromi the Control- Room in accordancee with the site Emfergencey Plan, and typically within 1 5 minaiutes. EAL #3 The "site specific lIs6t of coAmmunications methods" should incelude all conmmunications m~ethods used to performF initial em:ergency' notifications to) the NRC= as desc;r-ribed in; the Site Emfergency Plan. T-he listing- should incluide inistalled plant equipmenit anld componients, anid not items owned and mainitained by individuals. These 12' 1 4" "r o ; .. ....- ; T.:...... 41 --+/--: ,flXT~ ...---- ----,- *1 .+metheds me typteally the tie telephene lines, teated 1--meFgeney NEAH-teation ý-ý'stem ti-PiFsj teleplione kne and eemmeFeial ECL Asinmn ~AatibLutes: 3.1 .1kC 67 1 P a g e CA1 ECL: lei4 ALERT Initiating Condition: Loss of (Feeter esel4RCS [PWR] OF RPV [BR)IA. .inventory. Operating Mode Applicability: CoAld ...td.n. Refioeling 5 Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the Alei4-ALERT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Loss Of (F [PJsa]RCS or PV [B'JJr]) inventory as indicated by level less than (4e-.peei-fle'-evel*)32 ft. 9 inch (+ 6 inches above hot leg centerline). (2) a. .vesse.RCS [PWR] or- RPVl [B3R]) level cannot be monitored for 15 minutes or longer AND b. UNPLANNED ineFeaserise in (SITE SPECIFIC-SUAIP ANDWOR TANK) ANVANY of the followina suimp or tank levels in Table C2 1e-els-due to a loss of (reactor vessel/RCS [PpVR,] OF RPV[fBPR4,)inventory. Table C2: RCS Leakage* Containment Normal Sump" Pressurizer Relief Tank (PRT)* Reactor Coolant Drain Tank (RCDT)* MAB Sumps I thru 4* Containment Penetration Area Sump* SIS/CSS Purnp Compartment Sumnp Basis: This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.EAL #1- A lowering of water level below (site specific level) elevation 32'- 9" indicates that operator actions have not been successful in restoring and maintaining (reactor vessel/RCS [PWR] or RP%' [BWR]) water level.The heat-up rate of the coolant will inejeaserise as the available water inventory is reduced. A continuing deerease reduction in water level will lead to core uncovery. Although related, EAL #1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Heat Removal suction point). An ineieaserise in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.68 1 P ag e EAL #2- The inability to monitor (-reactor vessel/RCS [P,'VR, or RP, [BWR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/RCS. t ..F RPV{-B4,LRThe 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CSl If the (reactor vessel/RCS [424W] or ,. PW tr]) inventory level continues to lower, then escalation to Site Are Emer~gencySITE AREA EMERGENCY would be via IC CS].CAl: EAL-1 Selection Basis: The minimum RCS level at which an RHR pump can be started per OPOP02-RH-0001 is 32 feet 9 inches (+ 6 inches above hot ley centerline). If RCS inventory is reduced below this level, normal decay heat removal systems may not be available for core cooling. This threshold is not applicable to reduced inventory vacuum fill since this is a controlled evolution and not indicative of an RCS loss.CAI: EAL-2 Selection Basis: The tanks/sumps selected for this EAL were obtained from OPOP04-RC-0003. Excessive RCS Leakag;e. Since other system leaks could raise levels in various tanks and sumps. the list was limited to the tanks and sumps that would have the highest probability of indicating RCS leakage inside the Reactor Containment Building Although procedure OPOP04-RC-0003 is designated for use in modes 1-4. its logic is applicable to this EAL.

REFERENCES:

1. OPOP04-RC-0003, Rev. 18, Excessive RCS Leakage 2. OPOP02-RH-0001, Rev. 63, Residual Heat Removal System Operation Developer-Notes:2 Fer EAL6 # 1 the '"site specific leNvel" shouild be based en either: " [BU'R] Low~ Low EGGS actuation setpeint/Lcvel

2. This setpoint was choesen because it is a standard oprainally significant setpoint at whichi some (typically high pr-essure ECCS) injection systems would aui-toma-Ptically stait and is a value significantly below the low RPY wvatei lev~el RPS actuiation setpoint specifed il IC CU 1..[......] The minimum all.wable level that supports opertian ofiinormally used decay heat removal systems (e.g., Residual Heat Removal or- Shutdowxn Cooling).

if multiple levels exist, specify, eac-h along with the appropriate mode or- configuration dependency cr-iteria. For HAL 112 The type and range of RCS lev~el instr'umentation may vary, during an outage as the plant moves througgh various operating modes-, -and- refuieling evolutions, pariciular-ly 9fo a PAIR. As appropriate to the plant designi, alter-nate means of deteiiiininig RCS level are inistalled to assure that the ability to MoNitor level within the range required by eper-ating procedures

A411 noAt be interrulpted.

The instrum~entatio rag ncssary to 69 1 P a R e sbpa4implementation of opei-ating proceedir-es ithCodShutdown a Refueling modemabedfi-n (e.g., narroawer) thor. that r-equired during, modes higher- than Cold Shutdaown. Enter any "site speeific sump and/ofr tank" levels that could be expected to incr-ease if there wvere a less E)0inentorey (i.e., thie-lost iniventefy wouild eniter- the listed sump or tank).ECL Assignment Attr-ibutes:

39.1.2.B 70 1P a g e CA2 ECL:Ae-t ALERT Initiating Condition:

Loss of atIALL offsite and *tALL onsite AC power to emergency buses for 15 minutes or longer.Operating Mode Applicability: Cold Shutdo-wn. Rluelin5. 6, Deffieled"-ffwpk-EEmergency Action Levels: Note: The Emergency Director should declare the A-let4ALERT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Loss of ALL offsite AND ALL onsite AC Power to (site specific emergency buses). oIALL three 4160V AC ESF Busses for 15 minutes or longer.Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site r-ea Emergcne.y STSITE AREA EMERGENCY because of the incFeasedadditional time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emergency classificatia n leveI EMERGENCY CLASSIFICATION LEVEL would be via IC CSI or AS--RSI.CA2 -EAL-1 Selection Basis: N/A

REFERENCES:

1. OPOP04-AE-0001, Rev. 44. First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus 2. OPOP04-AE-0004, Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus 33. 0PSP03-EA-00,Rv

'3). O-0002, Rev. 32, ESF Power Availability

4. Drawing OOOOOEOAAAA.

Rev. 24. Single Line Diagram. Main One Line Diagram. Unit No. I & 2 71 1Page Deveopes-Notes!F-or a power source that has mualtiple ggenertatos, the EAL and/or Basis section should reflect the minimu numfber' Of operatingi generators necessary for that source to provide adequate powcr to an AG emergency bus.For eamiple, if a backuip pow'er so)ure is eEoInpriScd Of tWO -,enifrators (i.e., two 50%1o eapaeity generators siz~ed to feed 1 AC emergency bus), the EAL, and Basis section must speeify, that both gener-ators for- that sourcee ar-e operatig The "site spcfcemergency buses" are the buses fed by Off-Site Or em!ergency' AG pwer'e sourcees that supply power I-- the electrical distribuition system tIhat power-s SAFETY SYSTEMAAS. There is typically 1 emer-gency per train of SAFETY SYSTEMS.I; , l.-The EAL and/crF Basis section may speeiA' use of a non safety' related power- source provided that operation of this sourcee is controelled in acco-Ard-afce A-ith abno)rIml orF emerI-gency o~perating procedUreS. Or beyon)Id deSigH basis accident response guidelines (e.g., FLEX suppor-t guidelines). Such pb;Avei- sources should generally meet the"Alternate ac soeirce" definition pro'.ided in 10 CFR 50.2.At multi un~it stations', the &46s may credit compensatory5 measur-es that ar-e procedur-alized and canb implemenited wvithin 15 minultes. Consider capabilities suceh as poer~e source croass ties, "swing" genierators, Other-power, sourcees descrFibed in abnormal Or emei-geney, operating prOcedurFes. etc. Plants thiat have a proeedurfalized capability to supply offsite AC power to an affected unit via a cro)ss tie to a companion unit may cr-edit this poerAe source in the EAL provided that the planined cross tie strategy meets the requirem~ents of 10 C-FR 50.63.ECL. Ass~gignment Atr~ibutes: 3.1 .2.B I 721 P a g e CA3 ECL:A~ei4 ALERT Initiating Condition: Inability to maintain the plant in cold shutdown.Operating Mode Applicability: Cold S.hutden, .. Refueling56 Exa aple-Emergency Action Levels: (I or 2)Note: The Emergency Director should declare theA-lei4 ALERT promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.(1) UNPLANNED in RCS temperature to greater than (site specific Tec-nial Specifiatioen 0-00ld shutdoWn., temlpe r.at limit) 200 0 F (Tava) for greater than the duration specified in the-felewin tab4e-Table C3.Table C3: RCS Heat-up Duration Thresholds RCS Status Containment Closure Status Heat-up Duration Intact (but not at reduced inventory Not applicable 60 minutes*Not intact (or at reduced inventory Established 20 minutes*-P-4J4D) Not Established 0 minutes* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable. (2) UNPLANNED RCS pressure inei-easerise greater than (site specifie pi....u.i. Frading)iLO si .(This EAL does not apply during water-solid plant conditions.-fPW-R]) Basis: This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. EAL #1-The RCS Heat-up Duration Thresholds table addresses an ine-reaserise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation in PNNLRP). The 20-minute criterion was included to allow time for operator action to address the temperature inefeaserise. The RCS Heat-up Duration Thresholds table also addresses an inefeaserise in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature i.+ei-eserise without a substantial degradation in plant safety.73 PPaze Finally, in the case where there is af i-ei-easerise in RCS temperature, the RCS is not intact or is at reduced inventory [-P-z, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.EAL #2- provides a pressure-based indication of RCS heat-up.Escalation of the CS] or ARS.lAS-emergency classitication level EMERGENCY CLASSIFICATION LEVEL-RS]1.would be via IC CA3 -EAL-1 Selection Basis: Table C3 was adopted from NEI 99-01. Rev. 6. This EAL addresses the concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal. A number of phenomena such as pressurization, vortexing, steam generator U-tube draining, RCS level differences when operating at a mid-loop condition, decay heat removal system design, and level instrumentation problems can lead to conditions where decay heat removal is lost and core uncover can occur. NRC analyses show that there are sequences that can cause core uncovery in 15 to 20 minutes, and severe core damage within an hour after decay heat removal is lost. The allowed time frames are consistent with the guidance provided by Generic Letter 88-17 and believed to be conservative given that a low pressure containment barrier to fission product release is established. CA3 -EAL-2 Selection Basis: An UNPLANNED RCS pressure rise greater than 10 psig provides a pressure-based indication of RCS heat-up.The pressure change, per NEI 99-01 Rev. 6, is the lowest change in pressure that can be accurately determined using installed instrumentation, but not less than 10 psig.

REFERENCES:

I. Technical Specifications Table 1.2 (Mode. Temperature, Power, keff Table)Developer-Notes: For- EAL #1 Enter- the "site spccifie Tcchnical Spccification cold shutdown tempertatur limit" where indicated. The RCS5_ should be considered intact 9r noAt 4intAc inl _aPCcordancc With Site specific rieria.For F EAL #2 The "site specific pressur-e reading" should b.Pe the l-ow" est change in preSSure that Ca.. be accrately dctermined using installed intuctto.but not less than 10 sg For PAWRs, this 1G wnd its associated EA~s address the concerns raised by Generic Letter 88 17, Less fT Hewr Reiievf. ,A niumibcr ef phenefmena such as pr-eSSurization, voek&ing., steam gcncrater U tube draining, RCS lcaveal diffierfences w.hen oper-ating, at a mid leap condition, decay hecat removal system design, and leve instrumcntation pro~blems can lead to conditions whcrc decay heat r-emoval is lost and corFe uncovery can occur.NRC analyses show that there are sequence-s th-at can; c-ause core uncovcry in 15 to 20 minutes, and se~er-e coree damage w~ithin an hour after decay heat r-em oval is lost. The allowed tiffc frames are cosistent with the 74 1 P ag e guiidanee provYided b)y Genleric Letter Containmenit bairFier to fissieii produi ECL Assignment Aittribates: 3.1.2.BT I II 1 1 I D ,I .I 12 17 and behleve to b-e Consevtv ie that a IONS Sre--,.re et roicase is estaciisnee. 75 P a a e CA6 ECL: Alei- ALERT Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.Operating Mode Applicability: Cold Shutdown, Refiieling 5,6 Action Levels: (1) a. The occurrence of ANY of the following hazardous events in Table C5: Table C5: Hazardous Events* Seismic event (earthquake)" Internal or external flooding event* High winds or tornado strike" FIRE* EXPLOSION" (,:itespeitic .azards)Predicted or actual breach of Main Cooling Reservoir retaining dike along the North Wall" Other events with similar hazard characteristics as determined by the Shift Manager AND b. EITHER of the following:

1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.OR 4-2.The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.Basis: This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.EAL-#l.b.1-addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.

The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.76 1 P age EAL#-I .b.2 addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.Escalation of the emergeney CSI orARSIASIRS1. classification le-'el EMERGENCY CLASSIFICATION LEVEL would be via IC CA6: EAL-I Selection Basis: The listed hazards are taken directly from NEI 99-01, Rev. 6. The only additional hazard was the inclusion of the Main Coolin2 Reservoir since it is a credible hazard and analyzed in the STPEGS UFSAR (reference 2).

REFERENCES:

1. STPEGS U.FSAR, Rev. 13, Section 3.4. 1. Flood Protection Deve'loper Notes: For- (site _pecifi. hazar-d_).

develop OFS should ecnsidcrf ineluding ather significant, site specific hazar-ds to the Dulletee fist conipinco in LiL Ia ke.g., a setehe).N~eleaf PEAeFe plant SAFETY SYSTEMS eomprised of two or mor~e separate and r-edunfdant trains of.14 -AP-0-0-F-APP-3ea r-.+L-Q 1-31-46 W EleSlell eFlIeFia, LUkn, Lxz'sslonmnfii 'Lkrinates: -). I .-.77 1 P a g e CS1 ECL: SiteArea Emergen.y SITE SITE AREA EMERGENCY Initiating Condition: Loss of (eaeteFvesselIRCS tPWR] or RPo,, [BWR]) inventory affecting core decay heatremoval capability. Operating Mode Applicability: Cold Shutdown, Ref-ieling 56 E-4ample-Emergency Action Levels: (I or 2 or 3)Note: The Emergency Director should declare the Site Area Emergenty SITESITE AREA EMERGENCY promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.(1) a. CONTAINMENT CLOSURE not established. AND b. (.Reaetai-vese RCS [DWR] or RPV [BIR] level less than (site specif. c level) 33% ofplenum.(2) a. CONTAINMENT CLOSURE established. AND b. (Reae.taF ,essel,'RCS [PsWR] or- RP%1 [BWR]) l evel less than (site specific ievc!).(2) a. CONTAINMENT CLOSURE established. AND b. RCS level less than 0% of plenum (2) a. 3) a. [PWR] or RPWN [Br3ARo) level cannot be monitored for 30 minutes or longer.AND b. Core uncovery is indicated by ANY of the following: ,c......e specific r.adition ......) Reactor Containment Building. 68'-0" Area Radiation Monitors RE-8055 or RE-8099 reading greater than 9,000 mR/hr.OR:ý ..... pi ... u.)* Erratic source range monitor indication. ORWPWR4* UNPLANNED i-eFeaserise in (site specific suemp and/or tank) ANY of the following sump or tank levels in Table C2any, of the follo-ing OFnps or tank leNvels-of sufficient magnitude to indicate core uncovery.0 (Other- site specific indications) 78 1 P a y e Table C2: RCS Leakage* Containment Normal Sump" Pressurizer Relief Tank (PRT)* Reactor Coolant Drain Tank (RCDT)* MAB Sumps I thru 4* Containment Penetration Area Sump.SIS/CSS Pump Compartment Sump Basis: This IC addresses a significant and prolonged loss of (reactor vessel/RCS [PJM] Er RPM [BWR]) inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Ara S-gT-SITE AREA EMERGENCY declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor VesSel iev-eRCS level. If RCS/reactar vessel leveIRCS level cannot be restored, fuel damage is probable.Outage/shutdown contingency plans typieal4y provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCSreacto: -'essel leveWRCS levels of EALs I.b and 2.b reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. In EAL 3.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor (reaeter vesseWRCS [pUR] or RVl [BWR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level camnot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the ..eaetei..esse!/RCS [LR] .. r RPM [..1]." These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of/Shutdown and Low Power Risk Issues; NUREG-1449., Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guiclelines for Indust.y Actions to Assess Shutdown Management. 79 1 P a g e Escalation of the emergency classificaticn level EMERGENCY CLASSIFICATION LEVEL would be via IC CG 1 or-A-G-I-RG 1.CS1: EAL-1 Selection Basis: Per NEI 99-01 Rev. 6, the RCS level indication should be six inches (6") below the bottom inside diameter of the RCS loop penetration at the reactor vessel. Six inches (6") below the bottom inside diameter of the RCS hot leg nozzle (elevation 3 1 '-0.5") is elevation 30'-6.5" per OPOP03-ZG-0009, Mid-Loop Operation. Addendum 1, RCS/RHR Simplified Elevation Diagram. The nearest RVWL Monitoring System thermocouples are located 6 inches above (Sensor 6) and 4.9 inches below (Sensor7) the prescribed elevation of 30'-6.5". When water level is at the desired elevation of 30'-6.5", Sensor 6 will be dry and Sensor 7 will be wet. This condition corresponds to a reading of 33% of plenum per OPOP02- 11-0002, RVWL Monitoring System, Addendum 1, RVWL Sensor Elevations. CSl: EAL-2 Selection Basis: Per NEI 99-01 Rev. 6. the RCS level indication should be approximately the top of active fuel (TAF). The RCS level which corresponds to the top of the active fuel is 28'-2" (OPOP03-ZG-0009. Mid-Loop Operation. Addendum 1, RCS/RHR Simplified Elevation Diagram). The nearest Reactor Vessel Water Level Monitoring System thermocouple to TAF is Sensor 8 at elevation 29'-2.7". Use of RVWL to approximate TAF: with the inherent gap of 12 inches between indicated level and actual level, is acceptable for the purposes of signaling that the threat to the public is reduced when CONTAINMENT CLOSURE is established-whe*. CSI: EAL-3 Selection Basis: As RCS level drops the dose rates above the core will rise. Area Radiation Monitors RE-8055 and RE-8099 are located on the 68'-0" elevation of the reactor containment building. Their locations are identified on drawing 9C129A81105. Their range (0.1 mR/hr to 10,000 mR/hr) is identified in Table 12.3.4-1 of Section 12 of the UFSAR. A rising trend on these monitors can be an indication that core uncovery is occurring. Additionally, erratic source range monitor indications, or large level rises in the tanks listed can give further indication of core uncovery.The threshold value for radiation monitors RE-8055 and RE-8099 was based on Calculation STPNOCO 13-CALC-006 Rev. 1. The calculated monitor response is 22.4 R/hr when RCS level is at the top of the active fuel and 6 R/hr at one foot above the top of active fuel. The high range of these monitors is 10 R/hr. The value of 9.000 mR/hr was selected to ensure that the threshold is readily assessable and within the calibrated range of the monitor. The threshold value of 9,000 mR/hr corresponds to approximately 8 inches above the top of the active fuel with the reactor head on: which provides an additional indication that RCS levels are near the point of fuel uncovery. These monitor readings in coniunction with the other threshold values allow for an accurate assessment of the EAL.Core uncovery can be determined by the secondary indications listed in this EAL. The secondary indicators of inventory loss include a list of tanks/sumps found in OPOP04-RC-0003, Excessive RCS Leakage. Since other 80 1P agPe system leaks could raise levels in various tanks and sumps, the list has been limited to the tanks and sumps that would have the highest probability of indicating RCS leaka.ge inside the Reactor Containment.

REFERENCES:

1. Calculation No: STPNOC013 CALC-006 Rev.1. Dose Rate Evaluation of Reactor Vessel Water Levels during Reftieling for EAL Thresholds

2. OPOP03-ZG-0009, Rev. 59, Mid-Loop Operation, Addendum 1. RCS/RHR Simplified Elevation Diagram 3. USFAR, Rev. 15, Chapter 12, Table 12.3.4-1 4. OPOP02-II-0002, Rev. 15, RVWL Monitoring System 5. OPOP04-RC-0003, Rev 18, Excessive RCS Leakage 6. Drawing 9C129A81105, Re. 3, Radiation Zones. Reactor Containment Building.

Plan at E. 68' -0" DevelepeF NetesýAccident analyses suggest that fuel damage may occur within one hour of uncover~y depending upen tile amouint of timfe since shutdown; referF to Gener-ic Lelff 88 17, SECY 91 283. NUREG 14149 and NUMARC 91 06.The type and range of RCS level i.str.um.entati. n may Var rin. an. ou.tage as the plan.t mo..ves thr.ugh various operating modes and refuleling ev'olutionS, pailicularly for A PA/R. AS appro~priate to the plant designI alter.nate .eans f determining RCS level are installed to asse that the ability to monitor level withinl the rane rquiedby operating proEedureS Will not be interru~pted. Thke instrumentation range neessary--to support implementationi ofoper-ating, proceedur-es in the Cold Shutdown and Refuieling modes miay be different (e.g., na'rrwe"r) than that required dur'ing modes higher than Cold Shutdown.For EAL #l.b the "site spec.ific level" is 6" below the ..D of.the RCS loop. This is the level at 6" belo the bottom ID of the r-eactor vessel penetration and noet the low, point of the loop. If the availability of on scale level indication is suchi that this level value can be determined dur-ing some shutdowni moides Or conditions, but not others, thien specify, the mode dependent and/orF configuaration states dur~ingg which the level .Indcaio isf aplcbm h e ignad operation of water level inistrumenitationi is suchi that this lev'el value cannot be deter-mined at any, time durfing, Cold Shutdown or Refuieling, modes., theni do not include EAL#1 (classificatio n v;ill be accomplished in accordance with EAL #3).For EAL #2.hb The "site specific level" should be approximately the top of active fuel. if the availability of onl scale level indication is suchl that this level value can be deter-mined duringt, some shubtdov.n mofldes or-conditions, but not others, then specify the mode dependent and/orF coniguration states during which the leve inicaionis applicable. If the design and oper-ation of water level instrumentation is such that this level v~alue cannot be determnined at any time during4 Cold Shutdown or- Refudeling moHdes, then do not include EAL#2 (classifiation will be accomplished in accorFdancee with EAL #3).F-or EAL #3.b bu.llet As water level in the reactor vessel lowers, the dose rate above the core will increase. Enter- a "site specific. radHiation monitor" that could be used to) detect core unicovery and thie associated ';site specific value" indicative of core uneovery. it is reeognized that the condition described by this IC= may result in a r-adiation value beyond the oper-ating" orf display range of the installed radiatio monitor. In tho)se c-ases', EAL values should be@ dete#rmF~ined w~ith ýa mnargin sufficient to ensure that an accurfate; .. ..+; ; ............ .. ..... .. ..r .A 1, 811 Page monitor reading, is available. For- exiample, an EAL monitor- reading maighlt be set at 900% to 959% of the highest ac ..rate monitorF readin... .This no0t-'ithstanding if the esti.ated,;aleulated monitor reading is greater than approximately 110,0 of the highest acur0ate monior reading, th... d...p..S ma, choose not to include the monitor" as an indication and identifv an alternate EAL threshold. To furtl..her pr ...te accurate Classification, dev..Epers sho.I..ld con if SOme combinatiOll Of monitors cOuld b specified in the EAL to build in an appro~priate level ofecorroboration between monitor r-eadings into the clsiiatin assessm:enjt. For F=L 9.3.b second bullet Post T-MI acident studAies indieated that the installed PAIR instruamentation will operate errFatically when the core is uncoever-ed and that this should be used as a tool fcr making such determination+,s. For EAL #3.b thir-d bullet Enter- any 'site specific sump and/ori tank" levels that couild be expected to chanige if theare- wetre a less of RGSA-eete vessel iAnventory' of sufficient magnitude to inldicate corFe nIc-oVeryý. Specific level values may be incluided if desired.For EAL #3.b fourth bullet De. lps should deter... inle if other reliable in.dicators ex1st t. identify, fu.el...e+,e!y ,(e.g...... te viewing using, camer.as). The "oal is to identi6, any u e q!. site ,...il-.indications., not already, Lisedesv'e that ovill promote timely and accuirate emetrpency classific-ation+. BAIR Fo r SAL I-;. I.b "site specific level" is the Low Low Low EC-C- actuainson /Lvl1 h BAIR Low LEAS' LOW ECCS Aactuation-setpoint / Lev~el 1 was chosen becau'se it is a standard oper-ationally, significant setpoint at which some (typieally low.N pr-essur~e EGG(S) inj~ection systems would automati cally, start and attempt to restor.e RP.M level. This is A R.PV w Ievel value that is observable below the Low, , evel 2 value specified in IC CAl1, but significantly above the Top of Aetiv.e Fuel (TOAF) threshold specified i F=, L #2.For EAL #12.b The "site specific level" should be for the top of active fuel.For EAL; 43.b first bullet As wyater le'vel in the reactorF vessel lowers, the dose r-ate abee ~the e~wewill inceas. Enter a "site spec-ific radiation monito4r" thatcolbeudtoeetcrenoeyadth associated "site specific value" indicative of coeuevr.it is r~ecognized that the c-ondition described by'thjiS 1C mlay reul inA rad-iatio~n value beyond the oper-ating, Or display' ranlge ofthe inst-allead radiation monito. in those cases, EAL values should be determined with a margin sufficient to ensure that an accur.ate monitor r-ading is available. For example., an EAL monitor r.eading m.ight be set at 90%040 to 95(% o the highest accur'ate moinitor reading. This proevision notwithstanding. if the estimated/lealeuilated moitr reading is greater than approximately, 1101% of the highest accurate monitor Feading. then developers may choose not to include the monitor as an inidication and identify an alternate EAL thrfeshold. To further- pr-omote accrlate-Cl-ass'-ifi specified in the EAL to build in i o......... -n assessment., I I I Vcat'Aon, leveloper'ssoudcnie it soee comblinationl ot montors could be an aeofeeiriate level Ctcroorto etwleen onorre-adi~ners into the Pei- BWRs that do not bav.e installed radifatio n monitors capable 01 ind.i.atin core ..toeE)NeF', alternlate site speeitic level indicattons el core uneovery should be use d if available. 82 1 P a a e FEW EAL, 43.hb Seon-d- bulleIRt Becau-Mse BALR source r-ange moniter (SRA4) nuclear instr-umentation detector-S are typically loeated below corFe miid planie, thiS May- net be a Viable inidicato' Of ecore un~e6very fo0r B3As ForF EAL #t3.b th~ird bullet Enter any "site Specific SUMP anidOr tanik" le'VelS that cOUld be exEpected to change it- tfere wer~e a loss orf KFAr v mventeiry of sunil-eienm nagi may be included Of desir-ed.14HEle tO H iwameae core uncovery. Speemc level value s Forf E-,4 #3.b fcurth1 bullet Deve!opers should determinie if other reliable indicator-s exiist to idenitify, fuiel unever (.t. rmtviwing usinig camer-as). The goal is to identify any unique or site spec-ific-indications,. not already used elsewhiere, that will proemote timely and accurate emergency classification.A&ELL /\sstlzflmeft A"imrutes.: ~831 P a e CG1 ECL: Gener-al Emergency GENERAL EMERGENCY Initiating Condition: Loss of(reaeter- RCS [PW.] or inventory affecting ftiel clad integrity with containment challenged. Operating Mode Applicability: C-ld Shutdown, Refueling;.5 Exaompie-Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the General Em:crgeney GENERAL EMERGENCY-promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.(1) a. (ea.et.f ;esse. RCS [PR] or RPV [BTVR.) level less than (site specific leve4.0)0 of plenum for 30 minutes or longer.AND b. ANY indication from the Containment Challe:-ge Table C -e .(2) a. ReactOresse lARCS [RV..j or RP.. [B. j]) level cannot be monitored for 30 minutes or longer.AND b. Core uncovery is indicated by ANY of the following: (site specific radiation moenitor) Reactor Containment Building. 68'-0" Area Radiation Monitors RE-8055 or RE-8099 reading greater than 9.000 mR/hr.-s .p..ifi ae)..* Erratic source range monitor indication OR4f-4W-UNPLANNED iner-easerise in (SITE SPECIFIC SUMP AND/OR TANK) ANY of the following sumP or tank levels in Table C2 of sufficient magnitude to indicate core uncovery (Oth@r site specific inidicatiein

• C,;:tinme44:t

,.ts'imal S-:.n: o ~ STablee 12 tRu1* C,:teimAnr Pe:ctratic'; Area Sura AND 84 1 P a g e

c. ANY indication friom Table Cl the C:ntainment Chall T enge Table (see belb--W*.Table Cl: Containment Challene b-CONTAINMENT CLOSURE not established within 30 minuWtes,;

• ** >(Explosive mixture) 4% hydrogen exists inside containment
• UNPLANNED inefeaserise in containment pressure-WSeeandary, eontainment radiation moanitor I % r r-eaoing above (site speeitie value) [l~IF CONTAINMENT CLOSURE is re-established nrior to exceeding the 30-minute time limit. THEN declaration of a General Emergencv is not reauired.Table C2: RCS Leakage" Containment Normal Sniunp" Pressurizer Relief Tank (PRT)* Reactor Coolant Drain Tank (RCDT)* MAB Sumps I thru 4" Containment Penetration Area SLImp* SIS/CSSJPumiCtoa 1partmelt Stm ,1
• j , t It w, ,t a:IT LUIN+AJPOYIL.IN i LLUSUK is re estarviisnea priore to exceeoine the 3 u mfinute time limit. then aeeclaratien

-.I (9i .. ~ rr-.a-nA it kJCLiera f~C Ef iifelty kUL1N~t~'Lcf 1S II FequIbreUl. Basis: This IC addresses the inability to restore and maintain reaetef vesse!RCS level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel le-elRCS level. If RCS/reactor vessel levelRCS level cannot be restored, fuel damage is probable.With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency GENEP LGENERAL EMERGENCY is not required.85 1Pag e The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other liSted indications in Table C I to assess whether or not containment is challenged. In EAL 2.b, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor -(Feaetf-veseIRCS [PJR] er RP.. [B...) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reaetoE -esseILRCS [P;IR] or RPV [BW"R]).These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power. Risk Issues; NUREG- 1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and N UMARC 91-06, Gutidelines /br Industrv Actions to Assess Shutdown Management. CG1: EAL-1 Selection Basis: Per NEI 99-01 Rev. 6, the RCS level indication should be approximately the top of active fuel (TAF). The RCS level which corresponds to the top of the active fuel is 28'-2" (0POP03-ZG-0009, Mid-Loop Operation. Addendum 1. RCS/RHR Simplified Elevation Diagram). The nearest Reactor Vessel Water Level Monitoring System thermocouple to TAF is Sensor 8 at elevation 29'-2.7". Use of RVWL to approximate TAF; with the inherent Pgap of 12 inches between indicated level and actual level, is acceptable for the purposes of maintaining the escalation logic for the loss of RCS level condition. CG1: EAL-2 Selection Basis: The secondary indicators of inventory loss include a list of tanks/sumps found in OPOP04-RC-0003, Excessive RCS Leakagte. Since other system leaks could rise levels in various tanks and sumps. the list has been limited to the tanks and sumps that would have the highest probability of indicating RCS leakage inside the Reactor Containment Building.As RCS level drops the dose rates above the core will rise. Area Radiation Monitors RE-8055 and RE-8099 are located on the 68'-0" elevation of the reactor containment building. Their locations are identified on drawing 9C129A81105. Their range (0.1 mR/hr to 10.000 mR/hr) is identified in Table 12.3.4-1 of Section 12 of the 86 1Pave UFSAR. Rises on these monitors can be can be an indication that core uncover is occurring. Additionally, erratic source range monitor indications, or large level rises in the tanks listed can Pive further indication of core uncovery.The threshold value for radiation monitors RE-8055 and RE-8099 was based on Calculation STPNOCO13-CALC-006 Rev. 1. The calculated monitor response is 22.4 R/hr when RCS level is at the top of the active fuel and 6 R/hr at one foot above the top of active fuel. The high range of these monitors is 10 R/hr. The value of 9,000 mR/hr was selected for this threshold to ensure the threshold is readily assessable and within the calibrated rmnoe of the monitor_ The threshold v~ihle of 9.000 m R/hr with the reactor head on corresnondis to annroximatelv......... 8 inches above the top of the active fuiel which provides an additional indication that RCS levels are near the point of fuel uncovery. These monitor readings in coniunction with the other threshold values allow for an accurate assessment of the EAL.

REFERENCES:

1. Calculation No. STPNOC013-CALC-006 Rev. 1. Dose Rate Evaluation of Reactor Vessel Water Levels during Refieling for EAL Thresholds

2. OPOP03-ZG-0009.

Rev. 59, Mid-Loop Operations

3. Drawing 9C 129A81105, Rev. 3, Radiation Zones, Reactor Containment Building Plan at El. 68'-0" 4. USFAR, Rev. 15, Chapter 12, Table 12.3.4-1.

Area Radiation Monitors 5. OPOP05-EO-EO

10. Rev. 21. Loss of Reactor or Secondary Coolant 6. OPOP04-RC-0003, Rev. 18. Excessive RCS Leakage Developer Notes.!Aeccident analyses suggest + ftle, damage m..ay.o..r within one hc.ur of -no-very depending u.p.n t.. e rf time since shutdown; refer to Generic Lele88 17, SECY 91 283.. NUREG 1449 and NUMARC 9.1 06.The type and r-ange Of RCS leyvel inStrumentation may vary 'during anl outage as the plant moves throeugh variu-s-operating modes and refuieling evolutions, particularly for a PWR. As appropriate to the plant design, alternat.means of deterfmining RCS Ievel -are installed to assure that te ability to mo.nitor .level wl.ithin the rang-.e requir.ed byp perati..g prcedures will not be interrupted.

The in,,trum..entati on range necessary to sppoi, implementation of operatinlg pro~edurfes in the Celd Shutdown and Refttelinig modes mlay' be diffeent (e.g., nialfewer) than that r-equir-ed durfing, modes higgher than Cold Shutdown.For EAL 4 !.a The "site speifice level" should be app.r0wmately the top ofactive fuel. if the availability .f-n scale evl indic-ation is, su-ch that thiis level value can be determined during some shutdown moedes or conditions, but not others, then specify the mode dependent and/or configuration states during, wh~ich the level iniainis applicable. if the design and ope;ation of water- lev.el instrAmentation is such1 that this ...el v.alue canneo be determined at an .. time during, Cold Shutdown or Ref..eling m.des, then. do not include EAL .1 (classific.ation l b., 1he ac com plIshed, ,. inaccordance with EAL #2).For EAL #2.b first bullet As vwater level in the reactorf vessel lowes, the dose rate above the corFe will incerease. Enter a "site specific-radiation monitor" that could be used to deteet corFe uncover-y and the associated 871 Page retif; a aitinv lu'E bynd the operating, or display range of the installed radiation mon;itor. In thos)e caeSOS EAL values should be determined

with -A ri sufficienit to enisure that an aceurate moitOr4 reading. i-a-vailable.

Fr example, an EAL monitor Feading might be set at 900% to 95% ofthe highest aureaate mtnitor r-eading. This provision n .twithstandi.g, if.the estimated/calculated monitor reading iS -Feater than approeximately 110% of the highest accur-ate monitor- reading, then developers may choose not to include the monitor as an indication and idenitify, an alternate EA L thr-eshold. To furt~her promote accurate elarssificatiOnl, dleveloperS shoul1d cEon'siderf if Somfe combWHAinton o-fmonitors could b wpecified in the EAL to build in naprrate level of corroboration between monitor- readings inito the clasifcatonassessment. For BIARs that do fiot have installed radiationi moinitor-s capable of indicating core uneoer~ey, alternfate site specific level indications ofcore .nco ..ery should be used ifvavailable. For EAL #20h seoends bullet Post IMI accident studies iated t td... ...*.+; ...... k h ~alt as 1.1 ;.A; *,...d.e..

led that tik e Constalle WAI .....inistrumientation will operate erraticali5y when the cor-e is uincovered anid that this should be uised as a tool for makng ucdeerinaios.

BcueBRSreRage Monitor (SRN4) nuclear instruameantation; detectors arc typic-ally located below, core mid plane. this may not be a viable indicator of core uncovery forF BWA For- EAL #2.b thir-d buillet Enter any, "site- specific sunip and/or tank" lev~els that could be exýpected to chiange if there were a loss of inventory, Of suffic-ient magnitudle to indicate corFe uncover;y. Specific level values may be iincluded if desiredl.FEor EAL #12.b fourthi buillct Deve. peps shoulld determine ifothier realiab-lea indic-ators &eist to identify, fuel u1GNcover (e.g., remote viewn .sig.cmers) The goal is to idenitify any unique or site specific indications,no alr-eady used elsewhere, that will proimote timfely anid accurate emfergencyj classification.. For the Containment Challenge Table:.Site shutdown contingency plans typically provide fOr r-e establishing CONTAINMENT CLOSURE following a loss of R.CS heat r-emoval or inventei~y conitrl functions. For "Explosive mfixtur~e", developei-s myenter the m~inimum contaiinment atmospheric hydrogen concentr-ationi necessary5 to support a hydrogen burn (i.e., the lower deflagration limit). A concurr~ent containment exygen concentration may be included iffthe plant has this indication available in the Control Room.For B\'/Rs. the use of secondarj,' eontainmcnt radiation monitor-s shouild pirovide indication of increased release that my be indicative of a clhallenige to seondari.y containiment. EOP maximum safe values becauise these vausae fsl eeog The "site specific value" should be based on the nizable and have a defined basis.& az I1

• 1 A lm4L- /kssianment Attrioutes:

4. !.4.88 1 P a g e 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)ICS/EALS Table E-1: Recognition Category "E" Initiating Condition Matrix UNUSUAL EVENT E-HU1 Damage to a loaded cask CONFINEMENT BOUNDARY.Op. Modes: .-4-ALL 891 P a e E-HU1 ECL: Ntificatin of Unusua! Event UNUSUAL EVENT Initiating Condition:

Damage to a loaded cask CONFINEMENT BOUNDARY Operating Mode Applicability: A41ALL Emergency Action Levels: (1) Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than: (2 times the site specific cask speeie.. tehnical specification allowable radiation level) on the surface cf the spent fAel cask. : a. 60 mnreni/hr (,ganna + neutron) on the top surface of the spent fuel cask OR b. 600 mrern/hr (gamina + neutron) on the side surface of the spent fuel cask OR b. 7000 rnrem/hr (gamna + neutron) on the side surface of the transfer cask.Basis: This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.The existence of "damage" is determined by radiological survey. The values for this EAL are 2 times the Technical Specification allowable radiation levels. The technical specification multiple of"2 times", which is also used in Recognition Category RA IC AU-I-RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.Security-related events for ISFSIs are covered under ICs HUI and HAI.E-HU1 -EAL-1 Selection Basis: NEI 99-01 Rev.6 states that the dose rate limits are 2 times the Cask Technical Specification Limits. Section 5.3.2 of the "Certificate of Compliance No. 1032, Appendix A. Technical Specifications For The HI-STORM FW MPC Storage System", states: 90 1 P ajz e

5.3.4 Notwithstanding

the limits established in Section 5.3.3, the measured dose rates on a loaded OVERPACK or TR4NSFER CASK shall not exceed the followiing values: a. 30 mnrem/hr (gamma + neutron) on the top of the OVERPACK b. 300 mnrem/hr (gammna + neutron) on the side of the OVERPACK, excluding inlet and outlet ducts c. 3500 mnrem/hr (waima + neutron) on the side of the CASK

REFERENCES:

1. Certificate of Compliance no. 1032, Appendix A, Technical Specifications For The NI-STORM FW MPC Storage System. Section 5.3, Radiation Protection Program. 10 CFR 72.104, Criteria For Radioactive Materials In Effluents And Direct Radiation From An ISFSI or MRS 91 1Page 9 FISSION PRODUCT BARRIER ICS/EALS Table 9-F-1: Recognition Category "F" Initiating Condition Matrix ALERT FA1 A--yANY Loss or aiiyANY Potential Loss of either the Fuel Clad or RCS barrier.Op. AMfodes: P Opu6ati:i, Nat Sta:itf'y, tar-tmp. Hat4 S4Htdmiow 1,2,3,4 SITE AREA EMERGENCY FSJ Loss or Potential Loss of anyANY two barriers.Op. AMfodes: Poi'e"r Oper'atio, ot Sta& by,______ S ..tt up, f tJS, ,. ....w... 1,2,3,4 GENERAL EMERGENCY FG1 Loss of anyANY two barriers and Loss or Potential Loss of the third barrier.Op. Modes: Power Oper'atio.'a, ......S.. 'n at Shue,.wn.

1,2,3.4 See Tabic 9 F 2 for- BMIR EALs See Tanble 9 F 3 femr PMIR EALs Dev'eloper Note: The aElac-ent logic flow diagram is for uise BY.deve'aoeers and is noet r-eauircd for- site seecific inmlementatiain: how..ever, a site spec-ific scheme muist include some type of userF aid to facilitate timely and accurlate classification of fission product barrifr losses an/3ptnillsses. Suchl aids are typically, comprised of logic fie'w, diagrams. "scoring-". rit',ria or heekbox tyýpe mnatricies. The user- aid logic must be cositetwih ht of the.aEaeent diagr-am.92 1 P auý e De~velpep Notes 1. The legie used for these iniitiating, conditionlS reflects the following eensideratiens. _ The Fuel Clad Barrier and the RCS Ba..ier. are weighted more heavily than the Containment BRFr-e4-."Unusual Event ICs associated with fission pr-oduct barriers are addressed ini Reconiio Caegery S.2.- For aecident conditions involving a radiological release, evaluation of the fission proeduct barrFier thresholds will need to be per-formed in cojnto ihdose assessments to ensur~e corr-ect and timely escalation of the emiergency classification. For- example, an evalbuation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose a sse ssment may 0indieate that an EAL for- General Emergency lC AG I has been exedd.3. The fission product barrier thresholds speceified-within; A scemlfe are exipected to reflect plant specific design and operating-characteristics. This may r-equir~e that deveJeper-s reate different thresholds than these provided in th-e gen.eic g..n.. c.e.4.. Alte p tation methods for the Reognit-on CatcgorF'F WSalld fission product barrier.hre s eaceptable and include flaw char4s, block diagrams, and checklist type tables.Developers mu.st ens..re that the site sp method addresses all possible thresheold combinations and classification outcomes sho.n.. in t ...B.R or PWR EAL fission product barrier tables. The NRC staff considers the presenitation miethod o~f the Recognition Categorfy F information to be an impo ..ant use- aid and may request a change to a parti;.lar. proposed .method if, among other resns, the change iis niecessary to promonte consistency across the indust=y-.

5. As used in. this Recognition Category, the term RCS lea.kage encompasses notjust those types defined in Technical Specifications but also includes the loss of RCS mass to any) location inside containm...ent, a secondary side system (i.e.. PWR steam. .generator tube leakage), an inter.facin system, or outside of conainment.

The release f liquid oer steam mass from the RCS due to the as designed...pe ..d .per.ation of a relief valve is not considered to be RCS leakae.6. At the Site Area Emergency level, c-lassification deision makers should maintain cognizance o howfarR pesentcnditions Are from meeting a threshold that wouild require a General Emfergencey declar-ation. For exNample, if he Fuel Cl.a-d -And- RCS fissioni product barriers were both lost, then there should be frequtent assessmienits of conltainment radioac.tie inventory and inert.Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lotPh mrec ir-ector-wouild have more assurance that there wa oimdiate need to escalate to a Genieral Emiergency.

7. The ability to escalate to a higher emer-gency classification level in response to degrading conditions should be fmaiiatained.

For example, a steady incere-ase in RCS leakage would rpentan. icr-easmnv risk to public health and safet.93 1Page rT'a II l ll%'a * * .*Tam fi lo 'ip -- KV'K IC ai Uicuan *- iviur-t..a -'narr-a-r -AILR FS! SITE AREA EMERGN I FG1 GENERAL EMERGENCY A n' , Les oF an Potential Les.7 .7 a. Loss of Potentia eiteF he ietGla r RG bamrir I LOSS Ofall)'b. Less of any two barriers -an Less or Potential Less of the th*ld ba!--e!..F~iA CadBiiric dRCSýaRricr Coini nt Biarriei*f .OS POU-EN h-. LOSS i -POTEN LOSS POTENTIAL TIALTIALLOSS LOSS LOSS 4.P-~P*afI-YConta~ifnment Radiationt j.---4.-Pr-*fima*Containmcnt Radiation

k. 4.P-Pifffl~i, Containment Radiation A Pr;imar' ...ntainment

1. .. A. P,,imary ..ntainment ill Not f N A Pr..i.ay +ontainment radiatiOn m110ito01 Applieab radiatiOnl MOitor A-ppli-ea App~li~eab4 radiation 010nit40r reading gr-eater than le reading greater than 4ee reading greater than (site specific value). (Site-speeific value). (site-specific valae).Eo. 5.

p. 5. !icetions

5. 5. Other Idica.tion, A. (site specific as k. (site spe.ific as A. (site spe.ific as A. (site sp.eifi; as A. (site spe.ific as (site speiAfi, as ITF M7applieaeyeable a~- p~a~)p eabt4 i-. 6.... E Di..e.tor-judgment s. 6. Emcrgeney Director judgment t. 6. Emcr- ,y Drector Judgment A. ANYV conidition in A. ANNY conditionH inlth A. ,AN~t eeliditioIn in the A. ANYV conditioii in ,A. ANYV condition in the A. ANY/ conditiOnl in thee oiin afthe opinonefflie Oea-flil ffi he oinion a OeifEMefthe til~eopiniEiftl Of gmerg-@eny Director Emergency)

Dir-ector Emergencey Director Em~ergency Dir-ector Emergency Direector Em~ergency) Director'that indicates Loss thatindieates that indic-ates Loss of that ndieates that indicates Loss of ta niae of the Fuel Cla.Potential Loss ofthe the RCS B... Potential Loss of the the Co..ainment Potential Loss cf.the gafi-ef -Fuel Clad Barrier. RC+Ii.e -3A-FleiA -CItntainment Barrier.f-, 94 ....P.a ... Basis information For BWIR EAL Fission Product Barrier- Table 9 F 2 BMIR FUEL CLAD BARRIER TPHRESH4OLDS: The Fuel Clad barrieir consists of the zirealloy oi- stainless steel fuel bundle tubes that eontain the fulel pellets.1. RCS Activity This thireshold indi2ates that R. S radioacgtivity is greaer than 300 '-Ci/gm dsse equivalent " 13-1. Reaetor coolant activity above this level is 'greater than that expected for- iodine spikies and corresponds to an approimeinate r-angc of 290 to 51% ftuie clad damiage. Sinee this condition indicates that a significant amount of fuel clad damage has occur-red, it represents a loss of the Fuel Clad Barrier.There is no Potential Loss thrieshold associated with RCS Acti;ity.De~elepeF Noes;~Threshold values should be determined assuming RCS radioactivity concentr-ation equ.als 300 pti/gli dose equivalent I 13 1. Other site specific Emits may be used (e.g., tCi/cc).Depending uipon site specific-capabilities, this thrfeshold may, have a sample analysis c pnetad/or a radiatio mnitor readinig componenit. ~Add this paragraph (orF similar v'crE)ding-) to) the Basis if the thr~eshold includes a sample analysis component, "It isF recognpizied that sample collection and analysis of r-eactor-coolant with highly elevated activity levels, could requir-e sever-al hours to comsplete. Nonietheless, a sample related thrfeshold is included as a backuip to other indications." 2. RPV Water Level The Loss thrfeshold i'cpresents the EOP requirement for pray cotiment flooding. This is identified in the BsAROG EPGsiSAGs when the phrase, -Primaiyt Conitainmnent Flooding, is Requirfed,' appears. Since a site specific RPM water-level is not specified here, the Loss thr-eshold phrfase, "Primary) contaimnment floodin-g required," also accommodates the 90P need to flood the primar cot imet when RPW water level cannot be determinied and core damage d&ie to inadequate core cooling is believe-dA to be Potential Loss 2.A This water level corrfesponds to the top of the active fuiel and is used in the EOPs to indicate a chiallenlge to core eaeling.The RPV water- level thrfeshold es the same as RCS bar-rier Loss thr-eshold 2.A.. Thus, this thrfeshold inidicates a Potential Loss of the Fuel Clad barr-ier anid a Loss of the RCS barrier that appr'opriately' escalates the emer-gencey classification level to a Site Area Emergency. T-his thr~eshold is considered to be exeded wThen, as specified in the site specific EONs, RPV water- cannot be restored and maintained above the specified level following depressurizatio~n Of the RPW (either- manually. auitomatically or by failr~e of the RCS bairier) or when procedural guidance or- a lack of.low pressrUre .RPM. ii.e.ti.n s.ur. pi ..lu Emergency R42 dep.. ess.ization. EOPs allow.. the operator A w..ide.choice.of... in.c ..tion sources to consider when i-estering RPV water level to within pr-escr-ibed limits. E;OPs, also speeie', depressuriziationi of the RPNM in or-der- to facilitate.water level contro .with low pressure injection sources. in some events, elevated PM pressure may prevent rPestoation of RPe waterl ievel until pressure drops below the shutoff heads of.available iection. souces. Ther.efo.e, this Fuel Clad barrier Potential Loss is met only after- either:ý I) the RPMI has been depr-essur-ized. or- requir-ed emer-gency RPMl d4.....on hd giving the operator an oppo.ity to assess the .apabilit:' oflow pressre injection , to.. .resto RP. water. level or- 2) noe .' pressure RP systems arc, available, prseludinjg RPM depressurization ina f pt to minimize loss of RP'.' inventorm.'. The termi "cannoet be restored and mlainitained above" means the value of RPM water- level is noet able to be brought abovNe the specified limit (to~p of active fuiel). The dtriaoneursanevaluation of system per-formance and availability in relationi to the RPM water level value and trenld. A thr-eshold prescribing declaration when a thrfeshold value eaiunot be restor-ed and maintained above a specified limit does not i-eqair-e imfmediate actioni simply because the currFent valuie is below the top of activ~e fuel, but does not pei-mit exEtended oper-ation below the limfit; the thr-eshold muist be conHsidered reachied as soonl as it is appar-ent that the top of active fuel cannot be alhained.in high power ATWS/failure to scr-am events, EOPs may direct the operator-to deliberately lower RPMI water level to thle top of ac-tive futel in or-der- to r-educe i-eaetar power-. RPM water level is then controalled. between the top of active fuiel and4..... ............... ........... ... .... ... .............. ....... ............ ... .............. ...... ......... ...... .........., .. ......... ......... .. ................ ... ....... ...... .............. ........... ..,, 95 1Page the Minimum Steamf Cooling RPMl Waler Level (N4SC=RAl6). Althebugh suceh aetion is a chiallenge to core eealin'g and the Fuel Clad barrier, the immediate need to reduce r~eator* powNer is the higher- priority. For SUch even~ts, W~s SAS or !S!-5 w~ill dictate the Cnted fr emergency R lassio ation.Since the less of ability, to determine if adcquate corfe cooling. is being provided presents a significant chiallenge to the fuel clad barrier, a potential less of the feel clad barrFier is specified. Dev-eloper Notes:!Loss 2.A The phrfase, "Prifimary containment flooding requrired,"' should be modified to agree with the site specific EOP phrs exit fromf.. all .OPS and ent ry4to the SAGs (e.g., drywell flooding r-equir-ted, etc.).Potential Los 2 A The decision that "RP3. water level cannot be. detearmined" is directed by guidance given in the RPW3 water- level controlj ctins of f the e PBsi.3. Not Applicable (included forF numbering consistency between barrier- tables)41. Pr-imar~y Continmienit Radiation The radiation montor4 reading d to an instantaneous release Of all coolant massinton the p.. il.ar:.' .containiment, assuming that r'eactor coolanit activity equtals 300 EC-i/gmt: dose equiv~alent 1 13 1. Reaetor coolant actvi', above this level is gireater than that expested for iodinie spikes and corr~esponHds to anl aprxmaernge of 2-% to %fe cld dam~age. Since this condition indicates that a signiificant amouint of fue la amg has occurrFed, it represents a loss of the Fuel Clad Barrier.The radiation monitor reading in this threshold is higher than that specified for RCS Bar-rier Loss threshold 4I.A si 'nce it inidicates a loss of both the Fuel Clad Barrier and the RCS Barrier-.Note that a combination of the two monitor-radng appropriately, esc.alates the emergelncy. classification level to. a Site Area Emergency. T-here is no Potential Loss threshold assoeiatcd w.ith Primiarfy Containment Radiatiefon Developer Notes: The r-eading shou~ld be determnined assuming the instantanieous r-elease and dispersal of the reactor eoolant noble gas and iodine inv.nteiy, with RCS radioactivity concentration equal to 300 ..i.,/,m dose equivalent I 13 1, into the primary containment atmosphere.

5. Other indications Loss and/or- Potential Loss 5-.A This subeateggory addresses other site specific thr-esholds that may be included to indicate less or- potential loss of the Fuel Clad barr-ier-based on4 plant specific design chiaracteristics not considered in the generic gui danc-e.Developer Notes: Loss and/or Potential Loss 5-.A De~velopers shouild deter-mine if other reliable indicators exiist to evaluate the status of this fissionproduct barrier (e.g.r~eview, accident analyses described in the site Final Safety Analysis Repei4, as updated).

The goal is to identify a:~uiu of site spec-ific indications that will promonte timely and accurate assessment of barr-ier-status.Any. added thresholds should represenit approximately the same rel'ativ'e thrfeat to the barr-ier as the o~ther tki-esholds in this coIb:lumn Basis informationi for the other thr-esholds may be used to gauge the relati've bairrier thrfeat level.6. Emerg-ency Director juidgment Less-6.A This thrfeshold addresses any, ethcr facetors that are to be used by the Emfergeney Dir-ector in determiininig whether the Fuel Clad Barr-ier is lost.Potential Loass 6.A This thrfeshold addresses any, other facetors that may be uised by the Emer-gency Dir-ector in determinifing whether thle Fuel Clad Barrier is otnilyost. The Emergency Director-should also consider whethier or not to declare the bat:rier potntillylos inthe event that4 barrier sta~tus cannjot be moniitor-ed. Developer Notes: 96 1 P agR e NaE)The RCS Barrier is the reaeter coolant system pressure beundary and includes the RPM and all reaetor' coolant system piping ":p to and including the is-lation valves.1. Primary Containment Pressure The (site specific value) primar:y ..ntainment pr.essure iS !the dry', well high pr.essure setpoint Which indicates a LOCA by automatica.y initiating the or equivalent makeup System.There is no Potential Loss threshold associated -'Aith Primary Containment Pressure.De'veloper-Notes: Nolle-2. RPNI Water Level This water level cor-responds to the top of active fulel and is used in the BOPs to indicate challenge to core cooling.The RPM' water level thr~eshold is the same as Fuel Clad barrier Potential Less thr-eshold 2A~. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loess of the Fuel Clad barrier and that appropriately escalatesth emcgeney classification level to a Site Ar-ea Emergencey. This theshold is Aconsidered to be e..ceeded when, as sp.eifed in the site 912Os. RPM w .atec -annot be restor.ed and maintained above the specified level followinhg depressurization of the RP, (either mafnu.ally, automatically oF,-"b failure of the RCS barrFier) Or when pro)cedural gulidance or a lack of low pressurfe RPMP injection sources precluide Emergeney RPM depressurization EOPs allow the operator a wide of RPM injeetion sources to consider when restoring RPM water level to within pr.eseribed limflits. EOPs also specify depressut.ization of the RPM in order to facilitate RP ... .at. level -it "+'-. pressure i-jection som-rces. some events, elevated RPM pressure may prevent retoationi of RPM water- level uintil pr-essur-e drops belowN the shutoffheads of available injectioni soureS ThefetrBf. t RCS barrier Loss is me" o.n, _a4fe either:i 1) the RPMo has been depressu.ized. or required emergency RPM depfessurization has been attempted, giving the operateor an ,pp..unit. to ass" ss the capability of 1' pressure ijection sou.ces to resto. e RPM, w,.ater. level -or 2) low pressure RPM' iýection systems are available, precuding, RPD dep.essufization in an aeempt to i less of RPMI inventory. The termn, "cannot be restor-ed and maintained above," means the value of RPMV water- level is not able to be broHOght abov the specified limit (top of active fuel). The determinatio' requires an aluation ofsystemn perfomane and availability in relation to the RPM w.ater lvlvalue anid trenid. A threshold prescribing-declaration when a threshold value eaonno be restered and maintained above a specified limit does not require immediate action simply because the current value is belew the top of activ~e fuiel, but does noet permit etended opei-ation beyonid the lkimit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be alfained.in high po.eweFATWS/failure to seram events. EO12s may direct the oper-ator to deliberately lo~wer RPMN water level to the top of active fuiel in orFder to r-educe reactor power. RPM water- level is then contro~lled betwe~en the top of active futel and the Mnimutm Steam Coolin-g: RP. Water Level (MSCRAL). Although such action is a challenge to core coorin-g an-d the Fulel Clad barrier, the immediate need to r-educe reactor power, is the higher prk)ior'. For suchl events, W~s SAS orF SSS will dictate the need for emergency classifieation. There is no RCS Potential Loss threshold associated with RP,, Water Lev.el.RCS Leak Rate Loss Threshold .A, Large high;I energy lines that ruptuire outside primary).' cotainm.ent clan dischar.ge significant amounts of inv.entoery a jeopardize the pressure retaining capabilit,, of the RCS until they ar-e isolated. if it is determt~ined that the ruptured line cannot be p-ofptly isolated fom. the Contro;.+el Room, the RS ba r..ri. Loss threshold is et.Loss Thrieshold 9.B Emergency RPV Defressurization in accordance with the EOPs is i.dicative of a loss Of the RCS barrier. if Emergeoncy RPM Depressurization is per-formed, the plant oper'ator-s are directed to open safety relief valves (SR',s) and keep them+open. Eve. though the R. S is being vented into the suppression peoo, A Loss of the RCS barier eists due to the diminished effeeti~veness of the RCS to r-etain fiSSiOnl products Withini its bouHndary'. Potential Loss Threshold 3.A... .................................... ........... .....PI................................ ... .... .......................... .... ..................... ............... .......................... ...................................................... I. ......... .. ............... ... ..9 7 ..a ..e PDtential loss of RCS based on prim7ar system lcakage ..tside the primary c.. .tainm...nt is dt..mi.n. d from -EOP tem;peratrfe. o~r raito a omlOperating values in areas swell as main; steam line tunnel, RCIC, MPG!, oet.. which indiee~e a direct path from the RCS to areas outside primary containment. A~ Max Normal Operating valuie is the highest value o~f the identified pai-ametetr expected to occur dur-ing norfmal plant operating conditions with all dircctly associated aipport and contr-ol systems functioning properly'. The indicators rcaching, the thresh old ban'iefs and conffirmet-d-tc bca cauased by RCS leakage from a primfary' Sy'Stef Warranit anl A.e leiclassification. ,A primary system is defined to be the pipes. valves, and other- equipment which connect directly' to)the RPM suceh that a r-eduction in RPM pressur.e will ef4-et a dee..a.e in the steam or water being disehnged througha*n un~isal-atcd break in the systemi+.An UNISOLABLE leak whichi is indicated by M~a? Nonnal Operating valutes escalates to a Site Area Emer-gency w.hen coemb-inedd v.'ith CoAnt-Aimmeant Barriei- Loss thre-shold-3.A4 (after a coentainment iso-lation) and a Gener-al Emer-gency when the Fuel Clad Barrie:r criteria is also exeeded.DevelopeF Notes;Loss Threshold 3.The list of systems inceluded in this ~h~reshold Should be thle high eniergy linles Which, if ruIptured anld r-emiain unisolated, can rapidydpesrz the-a R W.. These- lIne are typically isolated by actuation of the Leak- Detection systm Large high energy, line br-eaks suchi as Main Steaim Line (MNSL), High Pr-essure Coolant linection (HPCI), Feedvater, Reactor- Water Cleanup (RIACAU), isolation Condenser- (IC) or- Reactor- Core isolation Cooling, (RCIC) that are UNISOLABLE represent a signifkicat los f the- RCS-& barrier-.4. Primary' Containment Radiation The r-adiation monnitor readinig corresponds to an inistantaneous release of all reactor coolant mass into the primaryL containment, assuming that reactor. coolant activity equals Technical Specification allowable limits. This value is lower than that spec-ified fcr Fuel Clad Barrier Loass thrfeshold 4.A since it indicates a less of the RCS Barrier only.There. s ano Pte-ntial Loss threshold associated With Pr-imarCnt Dcvclopcr Notcs: The r-eading shouild be determ~ined assumning, the instantaneouls release and dispersal of the. re-actor coolant noble gas anld iodinie invelntory, with RCS activity' at Technical Specification allowable limits, into the primary containment atmosphere. Using RCS activity at Technical Specification allowable limits aligsti hrsodwt IC S3 Alo RC-civt'a this level will typically' result in primiary containmenit r-adiation levels that can be mor-e readily' detected by primaryf)containmaent radiation moneiitors, and more r-eadily' diffeentiated fromi those cauised by pipg orcopoent "shinie" sources. if desired, a plant may use a lesser- value of RCS activity' for deter-minling this VAlue.in some cases, the site specific physical loc-ation and senisitivity' of the pri~mar',' contaminment radaton moitor(s) mfay' be such that r-adi-ation from -A cloud- of released RCS gases cannot be distingu~iShedp-from- rad-i-ationj emanajjtinjg from piping and components containing elevated reactor coolant activity. if so, refer to the Developer Guidance for Loss'Potential Loss&.A and deter-mine ifan alternate indication is av~ailable.

5. Other Indicationis Loss and/or Potential Loss 5.A This subcnte-geiy addresses other- site specific thrfesholds that may' be included to inidicate loss orF potenitial loss of the RCS barrier- based on plant specific design characteristics not consider-ed iii the generici guidance.Developer-Notes:2 Loss and'or Potential Loss 5-A Developers shouild- de-termine if other r-eliab-le indicaptor-s exis toevaluate thesatu of this fission proAduct barriear (.reve,' accidenit analyses described in the site Final Safety' Analy'sis Repoi4, as umpdated).

The goal is to identify' any uniu Or Site specific inictnstat Will pFromt tIml an acuate assess~ment o-f hbaarrier status.Any added thr~esholds should represenit approxiimately' the same r-elative thr-eat to the barriier-as the other- thr-esholds ini this column. Basis iifomiation forý the other- thr-esholds may' be used to gauge the relative barrfier threat lvl 6. Emerggency Dir-ector judgment Loss 6.A 98 1 P ag e This thi.eshold addresses an), efth.. f.. e...s that are to be used by the Emaer gency Director in deter..: ining Whether the RCS barrier iS 10st. Loss 6.A This threshold addr*esses any) .ther fat .rs that may be used by the Emer.gen. y Di-e.tor in detemining whethe. the RCS Barrier is p"tentially lost. The Eme..gen.y Dire.tcr-should also c.nsider ...ether 0!r not to deelare the barrier potentially lost in the event that barrier; stat. s .annot be m.nitor.ed. Developer Notes: Neti The Primary' ContAi;nmen Bar..ier inl.,des the diywel. , the wetwell, their respetive i...r.paths, and other connecetions up to and including the outermost containment isolation valves. Containment Barr-ier thresholds ar-e uiseda eriter'ia for- escalation of the E,-- fi..m Alek to a Site Area Emergency or a Gen.eal Emergency. 1.Pr-imnary Containiment Conditions Less 66t and 1AB Rapid UNPLANtED loss of primary containment pressure (i.e., not attributable to. dry.wel spray or condensation effects)following an initial pressure incr.ease indicates a less of primaRaiy co.ntainment in.tegrity. .Primary containment pressure should increase as a result of mass and e..ergy.. r elease into the primar-y .ontainment from a LOCA. ThRus, primar.y ecntainment press.re not increasingnder ... these conditions indicates a lass of pFrimary ..ntain....nt integrity. These thresholds rely on operator recognition of an unexpected r-esponse for the cond~itin -and therefore a spec-ific value i not assigned. The une.pected (UNPLANNED) r.esponse is impoi0ant because it is the indica;.tor for .a. tainment bypass eanditien. Potential Loss ] A The thr.eshold pressure is the primary e.ntain.men.t in.teral design pressure. Structur.al aeeptane. testing demonstrates. the c..pability of the primary contain.ent to resist presslres greater than the inte..al design pressure. A pressure of this and iseter than those expected to result from ay design basis aciden. t and, thus, represen.t a Potential Loss o the Containment barrier.Potential Loss I .if h!ydroge reaches or e..eeeds the lower flami.mability limit, as defined in plant EON-, in an oxygen rich envionmet, apotenitially exiplosiv~e m:ixture exists. if the com~bustible mixEture ignie inietepiaycnanet v~ ~ ~ ~ ~ ~ ~ ~~~~~~~~~~i .................................. ies inside the primary containinent, loss of the Containment barrie;" could occur.Potential Loss I .C The Heat Capacity Temperature Lim.it (l-IT. ) is the highest s.pp.ession pol temper.atue fo.m which Em.er.gen.y RP Depressurization will not raise: a Suppression chamber temperature above the maximum temperature eapability of the su.ppr-ession. cham..ber. and eipe w....ithin the su.ppression chamber which may be required to ope.ate when the R... is pressurized, OR 0 Suppr.ession chamber pressre above Prima.r.y Co.tain.ent Pressue Limit A, while the rate of eneg. transfer frm the R-PMl to the containment is greater, than the capacity of the conitainment vent.The HCTL is a function of RP;^ pressuire, suprsson.eel and suppression pool water level. it is utilized to preclude failure offthe containment and equipment in the containment necessary for the safe shutdowin of the planit an ther-efore, the inability to maintain plant par-ameter-s below the limit constitutes a potential loss of cntainiment. Developer-Notes: Potential Ls .BWR EPGs/SAgs specifically define the limaits associated with explosivea mixtuires, in terms of deflagration concentr-ations of.hydrogen and o....gen.. FOr ,1 I/W I .. ntainmen.ts the deflagration limits are "69; hydrogen. and 50/o oxygen in t dywell or suppression chamber". Fr-- N41- Ill e.ntainnents, the limit is the "Hydrogen Deflagration OverpressuHe t, .The thr.eshold term ..e-plosive mixtare" is synonymou.s with the EPG/SAG "deflagratikn limits".Ptential Loss !.C Sincee the HCTL6 is defined asuig age of suppression pool water levels as low as the elevation of the downeemef offenings in MLE 141 containiments, or- 2 feet above the elevation of the horizontal v~ents in a N&M H! cntakinment, it is unnecessary to consider separate Containment barrier Loss or Potential Less thr'esholds for abnormal suppression pool 99 1 P a g e water- level conditions. if desired, developers m~a:, incleude a separate Containmenit Potential LOSS thrfeShOld baSed an the inability to mnaintain suppression peal wvater level above the dov.ncomer openings in 4.1- 1111cotinens orf 2- fee~t abovea the elevation Of the horFiZontal venAts in -A MLIll W ontainmient with RPNV pr-eSSure abovNe the inmmdecay heat r-emoval pressur'e, if it will simplify Whe assessment of the preso pool level component of the HCTL.There is no Loss threshold associated with RPV Water- Level.Potential LOS'-2A The Potential Loss thr-eshold is idenitical to tile Fulel Clad Loss RPYl Water Level thr~eshold 2.A. The Potential Los 1-qi-i tfor Primar-y C01anetFlooding~ ind1cioaes dqaeCa- al cannot be restored and maintainled anld that core damage is possible. BAI'R EPGs/,kGSA sp~ecify, the conditions that r-equir-e primary containment flooding. When prima~' containment flooding is required, the EPGs are e-xited and SAGs are entered. Entry into SAGs is a logial escalation ini response to) the iniability to r-estore anld maintaini adeqtiate corFe ceoo 1g. PRA Studies indicate that the conditioni of this Potential Loss threshold eould be a cor~e mfelt sequence whichi, if not correceted, could lead to R-PN failure and increased potential forpimar onanment failure. in conjunction with the RPV water level Loss thresholds in the Fuel Clad and Cs barierclms this, threshold resuilts in the declaration of a Genieral Emergecy Devlper-Notes The phr-ase, "Primary conitainment floodinig r-equired," should be modified to agr~ee with tile site specific EOP phr~ase inidicating e~xit fr-om all EOPs and entry to tile SAGS (egdy Pelfooding-required, etc.).~.Pri-iar,' Containmient Isolation Failure Thes theshldsaddress incomiplete containment isolation that allows an UNISOLABLE direct release to the The use of the moditier "direct" iln defining the release path discriminates against release paths through interfacing, liquid mOFino-elease pathways, sul siit-iiftlines, Htprotected by thePifHFC-Htilel !eafi ytfi The existence o~f a filter is not considered in the thr-eshold assessment. Filters do not remove fision product noeble geases.in addition, a filter could becoeme ineffecti~ve due to iodinie and/or pa~iculate loading bey, on d designi limfits (ioe., retention ability has been exceeded) or wvater- satur-ationi froem steant'high hafiiidit' in the release streami.Following the leakage of RCS mass into prinlary containment anid a ris in p Imr containment pressur-e, there may b m~inrE radielegieal releases associated with allowable primary) containment leakage through various pefietrations or systemf compoenelts. Minorf releases may also occur- if a pi-iimar:, containment isolatiain valve(s) fails to close but thie pr-im~ary containment atmosphere escae to anenlsed system. These releases do not constitu'te a loss or* potential lo~ss Of primlaryýconitainmient but should be evaluated usin-g the Recogntitioni Categoriy A ics.EOPs may direct primary containiment isolation valv~e lo~gic(s) to be intentionally bypassed, even if offsite radioactivity r-elease rate limits will be exceeded. 914del thes@e onditions With a Valid primary) containmentOft i-sol1ationl Signal, thle cotainmenit should also be considered lo)st iff primary) containment venting is actually perfomied. Intentional venting of pr-imar',' containment for- pr-imary containment pressure or- combustible gas contro~l to the secondary containmnent and/or thle env~ironentlf.NA is -A Loss of the Containment. Venting for- pr-imary containment pressure conitrol when not inl an accident situation (e.g., to control1 pressure below the dryv.'~ell high pressur-e scr-am setpeinlt) does not mieet the thresheld condition. The Max Safe Oper-ating Teprtr n h a ae Operatinig R-adiation; Level Ave each; the higst value of tes p3aram~eters at whichi neither: (1) equipmient niecessary for the safe shubtdown of the plant will fail, nor (2) personniel access Ileeessar-y for- the safe shutdo'.n of the plant, will bepreluded. 9OAs utilize these temperatures and radiation4 levelst establshcod-itions unlder whichi RPV depr-essuriziationi is r-equir-ed. The temperatures and r-adiation levels should be confirimed to be caused by RCS leakage froam a pr-imar-y system. A primary system is defned to be the pipes, valves, aid other- equipment whichi connect directly to the RPV such thaa r-eduetiocn in PLPN pressur-e will effect a decrease i-n thteam or water beking dischar-ged thrfough anl unisolated breakE in tl eeomination witi RCS~ potential 3osA.A this tnresneiod wouldo result in a Site Ar~ea L~mergcnley. 100 1 P a g e There is ne Potential Les Developer Notes: Less 3.B 4rhes lie 1,A associated with Primai:' ..ntaimn.nt isolation Failure Consideration may be given to specifyinig the specific procedur-al step within the Primar-y Contaiimnment Control BOP that defines intentional venting of the Primar-y Containment reegardlcss of off-site r-adioactiv~ity release Fate_.4. Primary Cont.ain.ment Radiation Thkere is no Less thireshold associated with Primary Containment Radiation. Potential Loss '.A The r-adiation monitor- readinig corresponds to ani instantaneous release of all reactor coolant mass into the primaryL containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is .e.. abov tha. used to determine the analogoaus Fuel Clad arr4.ier Loss and RCS Barrier. Loss tlIphrehlds. NUREG 1223, Sew-ee Estima~hos Dtwiong lowidn: Responqfse ie Sevef-e Mieleao- Pomver ,Plont Aeekies, indicates the fuel clad failure mu.st be greater than approxim.ately. 20%.; in order for there to be a mjo.. r r.elease of radioactivity .equ.iring offsite pr-otective actions. For this condition to exist, there must alreadY' have been a less of the RCS Barrier and the Fuel Clad Bar.ier. it is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency' classification level to a Gener-al EmRergency'. Developer Nates: NUREG 1223, Seltrce Esýifncwiens During binident Response to Sei'eFre N::~erwP6.F P-ower- Plont Joeeots, provides tile basis for- usinig the 200% fuiel claddinig faýilure value. Unless ther-e is a site specific analy'sis juistifyinig a different v~alue, the reading should be deter-mined assuming, the instantaneous release and dlispersal of the reactor coolant noble gas and iodin inv.entefry associated with 2-0% fuiel clad failure into thlpimay contaMinment atmosphere. S. Other indications Loss and/0r Potential Loss 5.A This su.beategor' addresses other site specific thresholds that may be included to indicate loss or potential loss of the Containment bareier based on plant spec.ific d.si characteristics not considered in the generic guidance.Developer Notes: Loass ado Potential Loss &.A Developer-s shouild deter-mine if other r-eliable inidicatorFs exist to evaluate, the status of this fissionl pro~duct barrier e., review accidenlt analyses describ-ed in the site Fin-al S-afety' Analysis Reprt4, as updated). The go~al is to) idenitify' any!niu osite specific indicatioenss thapt will proemote timely' and accurate assessment Of ba~rrier status.Any' added thr-esholds should r-epr~esenit approxiimately' the samfe relative thr-eat to the barrier as the other- thr-esholds in this column. Basis information forf the other thresholds may be used to gauge the relative barrfier-threat level.6. E m ergency' Director-Judgmenit Less-6.A This threshold addresses any otIher facetors that are to be used by' the Emergency' Director-in determining whether the Containment barr-ier is losto Potential Less-".This threshold addresses any' o ther factors that may' be uised by' the Emerggency' Direector in deter-mining whether th Containmienit Baresptnillyelst. The Emergency' Diretor should also consider whethier or not to declare the barrier potenitially' lost in the event that barrier status cannot be monitor-ed. Developer-Notes: Noee 101 Page Table 9-F-23: p3AWR EAL Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FA1 ALERT FS1 SITE AREA EMERGENCY FG1 GENERAL EMERGENCY A-o-yANY Loss or a-+yANY Potential Loss of either Loss or Potential Loss of ai-yANY two Loss of aiANY two barriers and Loss or the Fuel Clad or RCS barrier, barriers. Potential Loss of the third barrier.Fueh"lad Barrierj. RCS...ariljer 2, Containmnent Barrieir: LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 1. RCS or SG Tube Leakage 1. RCS or SG Tube Leakage 1. RCS or SG Tube Leakage Not Applicable A. ,CS/reactor-ve*,,se A. An automatic or A. Operation of a standby A. A leaking or Not Applicable level less than (site manual ECCS (SI) charging (makeuttp RUPTURED SG Core actuation is required pump is required by is FAULTED Cooling -Orange ent ry by EITHER of the EITHER of the outside of conditions met following: following: containment. 1_. UNISOLABLE

1. UNISOLABLE RCS leakage RCS leakage VT OR OR 4-.2. SG tube 2. SG tube leakage.RUPTURE. OR aB. R.CS cocedew'n i'ate greater than (site SpecifiC prCeSuriZQel thlei-mal shock idieatieo-ns).

Integgity-Red entry conditions met 121 ~P age .E4 Fuel Clad Barrier. RCS Barrier (64 Containment Barrier (44) LOSS (4) POTENTIAL , LOSS (K4) POTENTIAL (-) LOSS (M) POTENTIAL LOSS LOSS LOSS (N) 2. Inadequate Heat Removal (-) 2. Inadequate Heat Removal (-P) 2. Inadequate Heat Removal A. @** ei A. C Not Applicable A Not Applicable A.. .spe.., feang&l+ : i-ei -aef eap-4iryhv-4ean into co..e -.oling temper-ature value). temperature vau, indicated by (, ite- P!Eeedwe-Core Cooling -Red Core Coolimn -.pecific id4icat,,4 ).... ... ........... ,.2.Restoratioi" entry conditions met Orange entry Heat Sink -Red .not conditions met entry conditions efFective within 1-OR met. minutes. Core B. i Cooling -Red entry heat re.!lo-al conditions met for heapa t v15 minutes orvia ctcamlog.I eat Sink -Red entry conditions met 3. RCS Activity / Containment Radiation Fuel 3. RCS Activity / Containment Radiation RCS 3. RCS Activity! Containment Radiation C-,d 'Ba2iefr BF*e B- iee Al. C Not Applicable A. Ce,+ti+-n'e'+4t Not Applicable Not Applicable Al. Gontaii'in,t r-adiation 1*monito faiatio)i' infltont.. radiationi mnitor redfi ae,!eadina 1greater i-eadin -eate~r t h m ( si t s ,. , 0 ......fl a.,(:eLpe i e-H--e+.-RCB Rad value). RCB R-a- a-e+l.RCB Rad Monitor RT-8050 oi;itoi. RT 84050 Monitor RT-8050 or RT-8051 efjTgýM or RT-8051 greater than- 40 .... +la 2 .gýreater than 380 R/hr R-*Not R/hr OR Applicable OR 2. HATCH 2. HATCH MONITOR MONITOR greater than 3-00 greater than 4-20 m R/hr 840 mR/lhr 1031 P age OR indieatlvs idates 4-34, Sample analvsis indicates that reactor coolant activity is ireater than 300 ItCi/gm dose equivalent I-131.104LPage S Fuel Clad Barrier : RCS Barrier Containment Barrier, LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 4. Containment Integrit or Bypass 4. Containment Integrity or Bypass 4. Containment Integrity or Bypass Not Applicable Not Applicable Not Applicable Not Applicable A. Containment A. Centainmiet isolation is required pi.es.... greater than AND EITHER of (ste .p.cii. ;ev.l,, the following: Containment -Red entry conditions met I. Containment OR integrity has been lost based B. Explosive mixture on Emergency exists inside Director containment judgment. (H,> 4%)OR OR C1. Containment

2. UNISOLABLE pressure greater pathway from than 9.5 psig. (site the containment speei... press to the setpoint)environment AND exists.2. Less than one full OR train of(&ie B. Indications of RCS .. .e ..... .leakage outside of e... ....containment.

Containment Spray is operating per design for 15 minutes or longer.105 1 P ag e Fuel Clad Ba.rier RCS Barrier Containment LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 5. Other Indications

5. Other Indications

5. Other Indications A. ei4e as A. (sie Sp AfA.e ..... e a, A. e,........

1.. A. -,....41.. ~ ~~- -1W a,,A .(site sp e ei.+' ass s.( .ite SF.e cifp e a " .. ... + 'appM4eab4e" N/A A applieable)-N/A [Aapplic-1;4.N/A app..eab.;leN/A "a p p.1...e. a -b.kN/A 6. Emergency Director Judgment 6. Emergency Director Judgment 6. Emergency Director Judgment A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the Emergency Director Emergency Director Emergency Director Emergency Director Emergency Director Emergency Director that indicates Loss that indicates that indicates Loss that indicates that indicates Loss that indicates of the Fuel Clad Potential Loss of the of the RCS Barrier. Potential Loss of the of the Containment Potential Loss of the Barrier. Fuel Clad Barrier. RCS Barrier. Barrier. Containment Barrier.106 I P a R e Basis Information For PWR-EAL Fission Product Barrier Table 9-F-23 Devecloper-Notes: Threshold Parameters and Values Ea.h WA.R owner's gir...p has developed a mctcdeloy f, guiding the developm.ent ad impleientation of Os(i.,aesigpatpaetrand determiining and prieritizing oaperateF ac~tions). Many of the thresholds coitained iii the PWR EAL Fission Product Bar'rier Table rFelect conditions that are spec-iflcally Addressed6 in EP__12 (e.,g., a less, of heat r-emoval eapability by the steamf gener'atorS). When developing a site specific threshold, developer-s shouild use thle parameter-S and values specified withinl their- EONs that alignl wit th1 cOW dto Ele::;ibed byý the generici threshold anid bas-is. adR1 late develaper notes. This approachl will enueensistenc btwe the site specific PONs and enierzeney C]',assifi-C',tion schleme. 'and' thuIs 4teci[tat'e timeb,) andJ aeeiiEc e::atccsisifieatioti assessnieflt'sem-ag-htemntatontheWestinghoeie-Arup(-)G) develepe4 a definedl set ofCriticeal Safet!y Funcltionis as par-t of their Emier-genc-Y Responise Guidelines. The WOG*pppr-al-h-tSuctures EOPs to we tnese resttiea! Safety R anto do so h:1-a pl-oritized anid systematic mianner. 'Hie M106 Critical Safe~ty Fuinetions al-c prFesentedl beON'.SSubce'itica!it:,' -Cre" Coo!ink RCS integrity'Con1tainmlent -RCS inventery The WOG- ERGs provide a methodology forf monitor-ing the status of the Cr-itical Safety Fu1nctions arid elassit'ilig thie signifi.an.e of a challen.ge to a f.netio...

this, is r.efer.ed toe Critical Saf~ety Fu4nction Status Trees (C-SFSTs).

Peo- plants that have imiplemented the WOG E=RGs, thle guidance i,901 allows for. use a CSFST assessm..ent ,esu.lts as EALs. and fissbion product barrfSeie l;ess/potentia 1O !H thi, Manner. an eer-gen,'y classifie.ation assessment miay fllodictly frm a CSFST assessm'ent. It is iMpor'tant to unlderStafrd that tile CSFSTs are-e ealuated uising plant par'aieter-s, andthat they ar f-pl-vend pe for puraposes of driving e....g.... .operating .......... usage. F _-. .... ..e.cnditions of inter.est, thle ..e..i tr1,esholds withi' flie PWR EAt Fis:fiy ,.. ... Baf--Table "e thle pant patmet+- ... .a-pý4&loss or loss of a f"is;o produFct barricr ho.. eve. as de.. .'ibed in tie associated Developer Notes, a C'F9ST ter'Imlinus m.ay lbe used as well. Por this -easn, inclusion. of thle C ..ST related thr..eshold would be redunfdanlt to thle paHrameter-based thr-esholds for- planits that emaploy the WAOO; ERGs.Sites that e1:Rp.Oy the W06 EROS ma... at theirE i,'clude the I(SST bas"ed losS and potential loss thr~esholds as descri-bled in, the Dev'eloper-Notes. Dev~eloper-s at these sites should consult with1 their-elassificationi decision makers to determiine if inclusion wiould assist Willh tiriiclvr and1 accurate effer-eenic' classificationi. This decision should consider-tile effects of aniy site sp!fiahngst the geri erie WOG CSFRST ev~aluation logic .aiad setpoints, as wyell as those arising fromi user rules appl ieable to enmetacne, oprtn ...procedures (e.g., e..cepti..ns to pr'eced're e..tr.y or- transitionl du" e to specific accident conditions E)r loss of a support S:,steff)." The .S.ST thresholds miay. be addre.ssed iioi oof 3 ways: 1. Net incor-porated; thr-esholds will uise par-amweter-s anld values as diseassed in the Developer Notes.2. incorporated a.lo.g With param.e.ter and value t-Ihesholds (e.g., a fAel clad loss would have.. 2 ti.resholds such as " T, "- 2000F" Cooling Red entr-y conlditions Met".3. Used in lie of par'arietet-s andI valules kfo all thresholds. With one exceeptieii. if a decision is mnade to incelude thie CSFST based thrfesholds. thien all uc alleowed.. .. ... ... .. ... .... .... ..... ........... ................. v _ .1111R I' attc -va thre..olds muti,1 be...... rsd in..... the table.i (e.., it is ...: permissible to use nl the C .1 t s as a 107 IP a ge poten-ial loss of the fuel elad d an.d diSregard all Oth... CSFST ba.Sed thr.eShOlds). The s.cpinis !he RCS integr1"ýFI-y I P) C-SST. Becaus-e 4f the copiexity ef the P Red deC-iSiOn po0int thatt I-Fe4iet lo,-,; t!STP i Guid Since breshOld WithouRt the- need to incorpOra'te the-eeher CSFST based threlsholds. s part of the Westingzhouse Owners Group (WOG) and has adopted the WOG Emeriency Response elines (ERG). These guidelines employ the use of Critical Safety Function Status Trees (CSFST).STP has implemented the WOG ERGs, the guidance in NEI 99-01 allows the use of certain CSFST assessment results as EALs and fission product barrier loss/potential loss thresholds. This approach allows consistency betwe en OPs and emergenvc classifications. 108 P a e -I-R FUEL CLAD BARRIER THRESHOLDS The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.I. RCS or SG Tube Leakage Loss 1 There is no Loss threshold associated with RCS or SG Tube Leakage.Potential Loss I.A in ed .ladding .... ag......Core Cooling -Orange entry conditions-met (CETs > 708' F) are sufficient to allow the onset of heat-induced cladding damage.2. Developer Notes:.Potential Las" l.A 41. Enterz the site specific r'eactor vessel wateir level valuEe(5) uised by E ON, to identify, a degr-aded core cooling condition (e.g., 'equires prompt reSteration action). The rcactor vessel lel that correspand to approx~imately, the top ofactive fuel may Also b e u-sed.5....or plants that have implentd Westighese Owner.s Group Emerigency Response Guidelines, enter tile reactor* vesse fevelks) used iorl tfie k. Ore 1-0014H kuranoe rain (inc6jluaigi aepenoelli u4S .pn the status of RCP~s, if applicable').

6. Westinghouse ERGC Plants ,7.2. Inadequate Heat Removal Loss 2.A This reading indicates temperatures

'.vithini tke core Core Cooling -Red en 1200" F) are sufficient to cause significant superheating of reactor coolant.try conditions me-t-CEl's >Potential Loss 2.A This reading indicates teiperatures Within tile Core Cooling -C Fjare sufficient to allow the onset of heat-induced cladding damage.)range entry conditions (E.Ts > 708'Potential Loss 2.B Heat Sink -Red entry conditions met (((NR level in all SG < 14% [34%] OR pressure in at least one S>1325 PS-G4SG)-AND total AFW flow to SG < 576 GPM). This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.-Meeting this threshold results in a Site Area EmergencySITE AREA EMERGENCY-_because this threshold is identical to RCS Barrier Potential Loss threshold 2.A; both will be met. This condition warrants a Site Area Emergency SITE AREA EMERGENCY declaration because inadequate RCS heat--109 Page PWR FUEL CLAD BARRIER THRESHOLDS removal may result in fuel heat-up sufficient to damage the cladding and ieieaeraise RCS pressure to the point where mass will be lost from thle system.I Develer Notcs:.e. ..specifie EON and/or EOP userguidelines ma' establish de.ision makiig cr.ite,,ia C onaefing the number or .therrt.fe.p .radig aeeesary tO drive aci;o... (;.g.. , reading gr.eater than 1 .200912 is required ber... g to an. inadequate c..re cooling procedure). To maintain eensisteney with EOPs', these decisioni irnaking 4riteria may be utsed in the. eeare exit therm,..c.,.uple

'eading thresholds.

La cc 2. A Enter a site speeific temper.atre value tat corr.esponds to signi..can. ill 00ore SUpe.eatilg of .eactog coolant. may also" be used.For plants that hav..e implemented Westinghouse Oners Group Emergenmy Response Guidelines, enter the parameters and values used in the Core Cooling Red Path.Potential Loss 2.A Enter a site specific temperature value that corresponds to cre. conditions at the on.set. of heat induced cladding da.. age (e.g.2hmru -hal fortespodsn foreation ofsuperheated steair assuming thathe RCS is intact). 7004' may al.so be used.For plants that have ime 1,mnted etihg house O wner.s ..oup Emergency Response GCidelines. enter the parameter.s and values u.sed in the Cor.e Cooling. Oran.ge Path.Potential Loss 2.R Enter the site speciic and values that define an. extr.em..e challen.ge to the ability to remove heat fi-om the RCS via thle steam generators. These will typically' be parameter-s and v, alues thai~t would requir-e oplerators to take prIp a."ion to addr-ess this conidition. For plants that have imnplemented We.tin'house Owners GCoup Response Guidelines. enter.the parameter-s and v~alues u~sed inl the Heat Sinkl Red Path.\Vestillhoulse E=RG Plants 6s a os inldication,. developerg shouild conisider-inceludinig a thr-eshold the. same as, or- similar- to. '-Core Coaling Red enltry coenditions met" inl accordance withi the. guidance at the fr-ont of this section.As a ShOUideeeh~~e e-o*ISidef nldn a flf-i-dtesm-s-tia o"Core Gealinig Oi-aige entry contditions, met" in accor-dancee with tile guidancc at the Pront ofthi; sectioni.As a potential loss inidication, developers should consider including a threshold the same -esr simia 4tor,"Heat Sinik Red cnitrx conidfitins wet", in accor-dance wit4h the. guidaiiee at thie frontii of fl4-see, 83. RCS Activity / Containment Radiation Loss 3.A. 1 The radiation readings for the containment high ranse area monitors (RT-8050 and RT-805 1) correspond monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment., assuming that reactor coolant activity equals 300[tCi/gin dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. The values for RT-8050 and RT-8051 were based on Calculation STPNOCO 13-004 Rev.2. The threshold values were conservatively rounded within 2% of the calculated values to make the values readily assessable. Temperature induced current (TIC)limitations are not applicable to the Fuel Clad Barrier Loss threshold 3.A.l because the expected radiation dose for this event overwhelms the TIC effect. This effect is discussed in the 10CFR50.59 evaluation 04-8245-60 associated with DCP 04-8245-33. .1"1.0. iiP0 ane... P-WR-FUEL CLAD BARRIER THRESHOLDS Loss 3.A.2 The HATCH MONITOR is located outside containment and is the back-LIp monitor to the containment high range monitors (RT-8050 and RT-805 I). The HATCH MONITOR threshold value is based on Calculation No. 03-ZE-003. This value corresponds to the calculated containment high range monitor readin.zs for Fuel Clad Barrier Loss 3.A The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold 3.A since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level EMERGENCY CLASSIFICATION LEVEL to a Site Area Emergency SITE AREA EMERGENCY. Loss 3.B This threshold indicates that RCS radioactivity concentration is greater than 300 ItCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred., it represents a loss of the Fuel Clad Barrier.Potential Loss 3.There is no Potential Loss threshold associated with RCS Activity / Containment Radiation. This mthr-eshold is the HATCGH MONITOR, whichi is the back uip monitor* to the containment high r-ange.monitors (RcT 8050 anid R-T 805 1) and is located outside conitainment aljacentt to the personnel ~hath it is recognized that sample collection and analysis of reactor coolant with highly' elevated activit;y evels c.uld reqUire sever.al hour-s to complete. Non.etheless, a sample rlted threshold is included as a backup to other iendcations. There is no potential loss tl--eshold associated with RCS Activity 'Contahnent Radiation. Developcr Notes: Tle Feading should be determined assuming tie instananeos release and.. dispersal of the reactoc ooelant noble gas and iodine inivenitory., with RCS radioactivity concentration equal to 300 ~t~i/igin doze equiv~alent 1 13,1, inito the otimn atmospher.e. Thr-eshold values shou~ld be deteki~iiind assuminig RCS, radio0activity3, eone entirati on equals 300 ptci~gfi dose euvenI13.Other site spec-tie units may be tused (e-g., Ptie)Depenidinig uponi site specific capabilities, this thres-hold may, have&asample nalysi comaponent and/or a r-adiationi moniitor-Feading componient.,Add this: p~aragraph (or- similarwding to the Basis if the threshold includes a sample anialysis componfenltý.1t is recognfized that sam~ple collectioni and anialysis of reactor- coolant with highly elevated no le ga .,<. ,, ; ..... ... .. .;a, , .' <' " '.,; ... ........w, ..1 -" q'a t ..... 4 0g 1 , .. ...included as a backup to Other. ii.dica.tionslý. .9-A. Containment Integrity or Bypass a a117P U M V IM" iv 11 111 lPage MR-FUEL CLAD BARRIER THRESHOLDS Not Applicable (included for numbering consistency) 4-40-5. Other Indications Loss and/or Potential Loss 5.A This subcategory addriesses Other Site specitic thresholdS that maly be icl-:ded to indicate lass or potential less of the Fu.. Clad barrier based en plant specific d en char-a.teriSticS nOt con.Sider.ed in !he gef-lei g-'idanee. N/A DeVeloper Notes: Developers should determine& if other- reliable inidieator-s exEist to ev~aluate the status of this fission proeduct barrier (e.g., r-eview accidenit analyses descrFibed in the site Final Safety Analysis Repor-t, as, upidated). The golis to identifY aHny unqeorstpecific indicationis that will proemote timely~ and accurate assessment ofbrirStatus. Any added thresholds should rersnfprxmtly te same relative threat lo the barr-ier as 4he other-.... : ......~~~.....................:.... '- 'V ........I'-......... a .. *:alTe..... .. ... .....barrier threat level.4-4-6. Emergency Director Judgment i 1,di the other tiisfitti Mak f~ u togE-et~ile rcii'e Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is lost.Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director-in determining whether the Fuel Clad Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. 112 1P a iz e IFWR-FUEL CLAD BARRIER THRESHOLDS Dceveoper-Notes: No~e 113 P P ... e PWR-RCS BARRIER THRESHOLDS The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.RCS or SG Tube Leakage Loss l.A This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage.It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location -inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment. A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of contai-nent, the declaration escalates to a Site Arca Em..geneySITE AREA EMERGENCY-_since the Containment Barrier Loss threshold I.A will also be met.Potential Loss l.A This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ECCS (SI)actuation has not occurred. The threshold is met when an operating procedure, or operating crew supervision, directs that a standby charging (makeup) pump be placed in service to restore and maintain pressurizer level.This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage.It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location -inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment. If a leaking steam generator is also FAULTED outside of containment., the declaration escalates to a s4e Afea EmeFg-eneySITE AREA EMERGENCY-since the Containment Barrier Loss threshold 1.A will also be met.Potential Loss 1..B Integritv -Red entry conditions This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock -a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized). 1)evelperNotes ,Aectaltien of the EGGS may alse be refcrrJe to as Safety. lIntieekn (SI) aetuatiein or other appiepi~iate site.Potential .ocs I.A 1141 Pa, e 4WR-RCS BARRIER THRESHOLDS Depe~nding upon char.ging ptl~m)tew.oum ... ...-pd+~44~eteF4el.epers-i4+ay-Use an RCS leak!rate valae of 50 gpmo a pporit site specifie v'alue. a-, anm altei-nate Potential Loss thr-eshold. Vf used, the thr-eshold wording shouild refleet 411a4 the dtrintc of the leak rate value excludes nor-mal redctien in RCSiventory (e.g.. b'. !he letdown systemn or RCP sea! lealzoff)-. Potential Loss 1.14 Eniter the site specific indications that define n;, eXtr-eme challenige to the integrit4y of the RCS piessur boundary dule to pressurized thermial shook a itransient that cauises r-apid RCS eooldown while the RCS is, in N4ode 3 OF higxher (i.e., hot anld pr-essur-ized). These ,vill typic-ally be parameter-s ansd valuesta WOUld reuie oprtr to take prompt action to addr-ess a pr~essurzized ther-mal shock condition,. Developers should also determine if the th-eshold needs to ,efleet any dependencies u.sed as r, trasitoneat.'decision points or- condition validation criiteria (e.g., an E011 used to respond to an excessive RC--od-own may not be enitered or- im:medi atelyv ex.,ited if RCS pressbtwe is bel ow a certain For- planits that have implemented Westinghouise Owner; Cr-oup EmerFeeaey Re-sponse Guidelines, enter, thle parameter.s and values u.sed in the R(-S i, v Red Path. Beeause of the comnplexity ofcertain, decision points within the Red Path of this C-SFST, developers at these plants ma., elect to not inelude the Specific par-amfeters and vaklues, anld instead follow the gUidAnCEQ ea.: W estinehOouse FRG Plants.As-a-potential loss indica"tion.. developers should consider-i..ek.din. a threshold the sami.e as, or. si.i.lar. to.noeted above , developer-s should en9sure that the threshold Word~inng refleets an', EFOP tran9isition1/enqtry decisioni points or- condition validation crziter-ia. For. exiample. a thrieshold might r-ead "RCs lIntegrfity (P)Red etycniin e ihRSpesr>30pi.

2. Inadequate Heat Removal Loss 2.A There is no Loss threshold associated with Inadequate Heat Removal.Potential Loss 2.A Heat Sink -Red entry conditions met (- NR level in allA," SG < 14% 134%] OR pressure in at least one SG) AND total AFW flow to SGs-SG < 576 GPM).This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.

Meeting this threshold results in a SiteArea SITE AREA EMERGENCY because this threshold is identical to Fuel Clad Barrier Potential Loss threshold 2.B; both will be met. This condition warrants a Site Area Emergency SITE AREA EMERGENCY declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and ifnereaseraise RCS pressure to the point where mass will be lost from the system.Devetop-rNaote-s+ lPotewitm l Los1 1151 Pa ge P3WR-RCS BARRIER THRESHOLDS Ent- hesie specific -e ...." ÷ h-b +- [ e t 1,q t- , ,i ;l +..... , .1. 1a. ... .... 1111 ...1.- " -1 .... ... " , , ; u, 11 .. .. .... .... .......... ....... .heat fr-om thle RGS via thle Steam gnror.Teewill typically be par'ameter's and valuos that wouEd reuie oeators to take prmp -cin to address. this econdition. Forplats hathav imlerontd Wstighouse Owýners Gr-oup Emergency Responise Guide.1ines eiie1: the parameters and valucs used in the !eat Sink Red Path.WestinhOulsc EpRG Pla nts HL,'k',e. ioee.is sheiiL;d e.iisi iLt meiLiiiii nE .tfifI S 10 8 t ie Sallie HS, er St L, f.. r i I miJlaf to. Heat 51flK med ei11,'f conditions met" in accordanee with the guH.3.. RCS Activity / Containment Radiation alG- e at- f~i E.;)RtE I 111 [i b' i;;.Loss 3.A.The radiation menitor- ieadingcalcul ation STPNOC-013 C;ALC= 004l provided a value- thaft corrI-esponds tO an instantaneous r-elease of all r-eactor cooelant miass into the conitainmnt asuig that reactor coclanit activity equals Tech.nical Spe.ification all..abvih.le lmits. The calculation resu lt..as that cspetainment f i-adiation motnitor-s wouild bredn'10mR'hr. These monitors have an overage bag-kground feading op 1- Wiffh dule to thel _9ene f a kep -lv"sore he vaRlue of- R1hr ;;As; peLectead bcuetha4t RCS. F80r ap dirox fratemy 40 minutoe fRlloyine a lp p.. roim l 40 mne , RT 8050 and RTreadings n texpe ete to .............. a apct -n ue ufei n h h-sel a i h udb i if a seeendf b" ...ak isd e I en-...e is ,susI,, ^e", eted.. .thi .... +, h e, .l l;I eShoul not used DtoQ 1..determine a eeoncurrent-l-oss of RC-S-frapoimt 90 m~inuteS.ThiS value is lower than that specifed for Fuel Clad Barrgier Loss threshold 3.A s~ince it indicates a loss of the RCS BarrFier only'. Not Applicable Potential [..oss 3.There is no Potential Loss threshold associated with RCS Activity / Containment Radiation. Developer Notes: The reading should be determined assmi':n',g thle in.stantanflee ws.release and disper.sal .f the reator. coo.lant noble gas and iodine inven.tory,, with R .S ,.tivit .at Tech.ical Specificalion, allowable limits, in'to thle containment atnfs4t-ere. U"sin:g RC-Saetkity at Technical l allowable limits aligns this threshold w.ith IC S-43. l I , R.CS activity at this, level will typically, result iin conitainmenit r-adiationi lev~els that can be mor-e readily. detected by containment radiation monitor-s, and mor-e r-eadily differ~entiated from those caused b:, piping, or component -shine" sources. if desired, a planit may' use a lesser- value of RCS activity for- detei-fininig thiS value.fin Some oases, the site speeii. hyia-lcto anfd sesiivtyofhe containmenit radiation monitor(s) may be such that r-adiation fi-om a clouid o4freleased RCS gases cannot be distinguished fi-om iradkition 1161 Page PIWRRCS BARRIER THRESHOLDS Developer Notes, 44 LtePgtenyial Bpss 4. Containment Integrity or Bypass s 5 An dppvmnpH+RiwA*e4 Ac4(w-eee4 F Not Applicable (included for numbering consistency)

5. Other Indications Loss and/or Potential Loss 5.A T-his Subeategoryý addresses other site specific thr~esholds that may, be ineluided to iidicatc l05s orF potenltial less of the RCS barriier based en plant speeifie design eharacter-isties not c-nsidefr in the gen.eric guidance.

N/A Dcvelopcr Notes: Los-, and/or- Potential Loss 5-6, Developer-s shouild deterinilfe if other- reliable indicators extist to c'.akiate the status of this Fission producet harri.e" (e.g.. re\','".. acden a....es des.ibed in; l the, ,ite inal, ysi Fer. as updated). 'h41e gEoal iS to idefitif,' any' un ~e0.stpeeific-inidieations that will pRo~note timely anid accurate as'sessmnent of barrier stats.Any' added thresholds should r-epresenit appro.ximately the same r-elative thveat io the-4+arries-as the other thresholds in this column. Basis infor-ma4ion fori the other thr-esholds, may be used to gauge. the r-elaiive barrier threat leve-.Variables used to monitor for the significant breach or the potential significant breach of fiel clad, the RCS pressure boundary, or the reactor Containment, are desigaated Type C.The response characteristics of Type C information display channels allow the control room operator to detect conditions indicative of significant failure of any of the three fission product barriers or the potential for significant failure of these barTiers. Although variables selected to fulfill Type C functions may rapidly approach the values that indicate an actual significant failure, it is the final steady-state value reached that is important. Therefore, a high degree of accuracy and a rapid response time are not necessary for Type C information display channels. Type C variables are found in UFSAR Table 7B.6-1.6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost.Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. 1171 Page I PW1-R-CONTAINMENT BARRIER THRESHOLDS D~oper- Notes!NetieThe Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the ECL from-e-4--ALERT to a Site Area Emergency SITE AREA EMERGENCY or a General Emergency GENERAL EMERGENCY. I1. RCS or SG Tube Leakage Loss ILA This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss I.A and Loss 1.A, respectively. This condition represents a bypass of the containment barrier.FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably [part of the FAULTED definition] and the fati--ed-FAU.LTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU3 for the fuel clad barrier (i.e., RCS activity values) and IC SU4 for the RCS barrier (i.e., RCS leak rate values).This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in a significant and sustained release of radioactive steam to the enviromnent (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment. Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold. Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category ArR ICs.1181 Page PWR-CONTAINMENT BARRIER THRESHOLDS The emergency classification levels-EMERGENCY CLASSIFICATION LEVELS resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.Affected SG is FAULTED Outside of Containment? P-to-S Leak Rate Yes No Less than or equal to 25 gpm (a,,l-hal+*-,er C1 Developer Notes)Greater than 25 gpm (-o other value per SUt4 De;'clOpfrf otes Requires operation of a standby charging pump (RCS Barrier Potential Loss)Requires an automatic or manual ECCS (SI)actuation (RCS Barrier Loss)No classification U+usual Event UNUSUAL EVENT per SU4 Site Area Emergency SITE AREA EMERGENCY per FS1 Site Area Emer'gen:cy SITE AREA EMERGENCY per FS1 No classification Unisual Event UNUSUAL EVENT per SU4 Ale4 ALERT per FA I Ale4 ALERT per FA I Potential Loss I.There is no Potential Loss threshold associated with RCS or SG Tube Leakage.Developer Notes;A.' team .....at. poer oper.ated r .elief valve may; also be r.efe.r.ed to as an atmnospheric steam dump valve ..r..ter.a.popr... " i. e specific ......Developer-S may include anl additional site 'specilic thre-Sho0ld(s') to at-ddress proelonged steami releases necessitated by oper'ational conisider-ations ifANOPS 8r FOPS cou1ld Fectblile that a leakingeiorRUPTPURE-D steami genefatef-be utsed to support plant eooldevwn. eveopers tabe intOu tSer. aids o. vallb 4 er loea4ions

v wiin theirh asi docum t.2. Inadequate Heat Removal Loss 2 There is no Loss threshold associated with Inadequate Heat Removal.119 19P a a e PAIR CONTAINMENT BARRIER THRESHOLDS Potential Loss 2.A Core Cooling -Red entry conditions met for 15 minutes or longer. This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and a++incrieasedehigie potential for containment failure. For this condition to occur there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.The restoration procedure is considered "effective" if core exit thennocouple readings are decreasing and/or if reaetor vessel vRCS level is increasing.

Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Director should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective. Severe accident analyses (e.g., NUREG-1 150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence. Notes: Some site specific E01 S BAnd/r EOP user- gidelinies miay establhsh decisionl miaking cr'iteria cnerneeiing the numliber OF Other attributeS of thermocEOUp~le readings, neee-Sary, to drive actions (e.g., 55 CETs reading-gr.ate.. than. 1,20O ..F is r-equired .re t-.ansiti.in. to an inadeq..a.. re e. ling pro.edu..e). TP&maintaini consistency with EOl. tese decision m~aking, crFiteria may be used in the cor-e e.xit thermocupe readi thresholds. Potential LaOss 2.A. 1 PEnter site specific er-itria. requir-i-g;3ntry kinte a core cooling restor-ation Proedurc OF pOrompt implementation 4oFre cooling r-estorationR acetionS. A rfeading of 1.200oF oni the CET-s may also be uised.Fo- planits that have implemiented Wetigoe Ower GrI Emrec espoiise Guidelines, enter the parameters and values uised ini the Cor-e Coolingf Red Path.wes tin zh use ERG Plant4 Dev~!)eloeS shouE!dld osiderinci!luding a thr-eshold the same as. Or Simfi!ar-to, "Care Coo! ling Red enitry conlditions Met for 1 H! minutesi OF longer" ini accor-dace withi the guidaniee at the front of this section.1201 Page PWR-CONTAINMENT BARRIER THRESHOLDS

3. RCS Activity / Containment Radiation Loss 3 There is no Loss threshold associated with RCS Activity / Containment Radiation.

Potential Loss 3.A. I The radiation readings for the containment high range area monitors (RT-8050 and RT-8051) correspond monitor reading eorresponds to an instantaneous release of aEll eaeeor coolat Rs-the radioactive material inventory of the reactor coolant syst.e (i.e.. All the RCS coolant mass) into the containment, assuming that 20% of the fuel cladding has failed. The values for RT-8050 and RT-8051 were based on Calculation No. STPNOC013-004 Rev.2. The threshold values used were conservatively rounded within 2% of the calculated values to ensure the values were readily assessable. This level of assumed fuel clad failure is well abe-,e-bevond that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds. Temperature induced current (TIC) limitations are not applicable to the Containment Barrier Potential Loss threshold 3.A.1 because the expected radiation dose for this event overwhelms the TIC effect. This effect is discussed in 1OCFR50.59 evaluation 04-8245-60 associated with DCP 04m8245-33.NUREG- 1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level EMERGENCY CLASSIFICATION LEVEL to a General Emergency GENERAL EMERGENCY. Potential Loss 3.A.2 The HATCH MONITOR is located outside containment and is the back-up monitor to the containment high range monitors (RT-8050 and RT-805 1). The HATCH MONITOR threshold value is based on Calculation No. 03-ZE-003. This value corresponds to the calculated containment high rangc monitor readinas for Containment Barrier Threshold Potential Loss 3.A. 1.Deveoper-Notes: Potential LOss 3.A N UREG 12_. So,'cc L:otic:,; Dr"'ing- Responsef ,qiq,5e eXe I -ower P!kc.ni Ieeiedk;i-s p'o"vides hasik fior using the 20%", fuel eladdi.-g failure value. tnlessherei. s a site specific analysis'ust-if a diff.er-ent value., the r-eadiiig gho.ild be deter-mined assu.. ing the in.t.a...ane.us release and 4mspe:sa! fhe ireactoir ec4,+t--ntb1ý-, s--mid4od-*e fiventor'y a.sociated h- 4fuebadfai4-we-i-nt& --e con.a.nment almospnere.

4. Containment Integrity or Bypass Loss 4.A 121 IPa,-e MIWRCONTAINMENT BARRIER THRESHOLDS These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both thresholds 4.A.1 and 4.A.2.4.A. I -Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage).

Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency Director will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).Refer to the middle piping run of Figure 9-F-34. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment. Following the leakage of RCS mass into containment and a rise in containment pressure., there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category AR ICs.4.A.2 -Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g., through discharge of a ventilation system or atmospheric leakage).Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.Refer to the top piping run of Figure 9-F-344 in Addendum 3. Containment Integzrity or Bypass Examples. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment. The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.Leakage between two interfacing liquid systems, by itself, does not meet this threshold. I 1221 P a e PWR-CONTAINMENT BARRIER THRESHOLDS Refer to the bottom piping run of Figure 9-F-43. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage friom the clesed water c.eling Component Cooling Water system to the Auxiliary Building, then no threshold has been met. If the pump or system piping developed a leak that allowed steam/water to enter the Auxiliary Building, then threshold 4.B would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A. I to be met as well.Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s)fails to close but the containment atmosphere escapes to a closed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category AR ICs.The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold l.A.Loss 4.B Containment sump, temperature, pressure and/or radiation levels will inef-easerise if reactor coolant mass is leaking into the containment. If these parameters have not i'ereasedrisen, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). ie-reaseRises in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment. Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be con'oborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not -neFeaserise significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc. should be sufficient to determine if RCS mass is being lost outside of the containment. Refer to the middle piping run of Figure 9-F-3-14 in Addendum 3., Containment Integrity Or B','pass.a.pl.es. fn this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A.I to be met as well.To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold I.A to be met.Potential Loss 4.A Containment -Red entry conditions met (containment pressure > 56.5 tSIG). If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a Site Aiea 123 1P a g e PWR CONTAINMENT BARRIER THRESHOLDS Emergeney -SITE AREA EMERGENCY and General Emergency GENERAL EMERGENCY since there is now a potential to lose the third barrier.Potential Loss 4.B The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limitI4!4.}). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of contaim-nent integrity. It therefore represents a potential loss of the Containment Barrier.Potential Loss 4.C This threshold describes a condition where containment pressure is greater than the setpoint (9j PSIG.P-P-4) at which containment encrgy,.heat) removal systems are Containment Spray is designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that ,ntainiment heat yreoval/depressurizatio, systems (e.g., con.tainment sprays, ice .onden.sei. fa.s... et.., bEi,...t in.luingo ..i........... t .ez. ies) are Containment Spray is either lost or performing in a degraded manner.Develeper-Notes: Leoss 'i.A.Developers mi.ay .e a liSt Of Sit. pecific ...di..tion .mnit.S to beter d.efine thiS. th-e.. ld..E.. ec.ed MO94ito1r alrmS OF readings may also be included.Potential Less !.,k The Site speci Pic press-re is the contain ment design pressure.Fer plants that have~iimplementied Mlestinighouse Owniers Cr-oup Emer-gencey Rcspons~e Cuidelines. tile pressure value in Potential Inoss 4.A is that used for- the Containment Red Path. if the Conitainment CPSS4 at of moron-etain t+ha-e, enter" the higl.est, c.ntainmi value Shown) on, the tr.ee. Thi' is typicall, the c.n.ainment Pr~es~sure. Potenitial Lkss 4.13 D)eveloper-s may entier the i:Hnunum ceintainimenit atmospherici hydr~ogeni concentr-ation neeessaryto, support a hydrogenl bur (i.e.. , l, .he.lo.e de.lag.a.ioni liit). A concu..ent contaiinm:.ent. con.entrati.. may' be incuded if thle Plant has t.is in.dication available in thle Controln Room.Potential Lesos~ 4.G.Enter the site specific pressurze seipoint value thiat actuiates containment pr'essur~e contr~ol systems, (e.g., cont~ainmient spr~ay). Also enter the cite specifiecotiet prsueentrol system/equipm~ent-tha should b- .pe' designi the containment press-ure setpoi,," is r. .,I.,,.. 1. desired. specific conditioni indEicatios5 51u10h as paramete-N'alues cani also be enter-ed (e.g., a containment4 spray h1ow r-ate less thian a certain Value)-.This threshold is not applicable to thle U.S. Eo l-tioniary. POWe' ReacOr (E)! dsie,.WestinAehouse ERG Plants, As a potential loss indication,. developers, shouild consider ineluding. a thrfeshold the sa~me as. or- simfilfr to.5. Other Indications cflmm;;I M, mi~ LaCreIaoi 2Ce 3Aitil tile '2L1UidaiCC atL tile K1r11[ Bi- [[IS :;eet!1.Loss and/or Potential Loss 5.A 124 1Pa e PW1 CONTAINMENT BARRIER THRESHOLDS This su,1bcat-g.oy addr-esses other site specitic thesh'-lds that ma;,' be-included to indicate lo,'ss ort peM+t-Ua 10S9 Of the COntainmen~lt barrier basied an plant specific desig n chairacteristics niat coansider~ed iii the genler-i guidance. N!A Devlper- Note I oss and/or Potential L.,oss 5.A 6. seemergency iretnrpovide f venting othe containment as a means of prevtenting aTaisthreshid ailureaLsse a threrhl fcosthatl ay included for the otimerncy bDriretr. Tin dtrerhoiningl be met as soeo as such venting is loINsENT. Containment ventin as part of rec-er s i7 classified in accor-dance with the i-adiolegieaj effliuent !Cs.Dev~elopers should dletermine if other- reliable indicators, exist to evaluate the status of !his fission product barrier. (e.g., review. accident analyses descr-ibed in thiesite.Final Sft'AnlisRepor

t. as updated).

The goal is te identifal any unie or site specific indications that will pr6mote timel and accurate assessment of barrier status.AHN' added thr-eshodS Should! repre.sent approximately the samie r-elative thrfeat to the harriier, as the othier thrshld inths olun.BaiS in41-foration for1 the Other threshOlds may' be us-ed to gauge the@ relative bar-rier threa.: level.6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost.Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Developer-Notes: Noee 125 1Pa e PAWR-CONTAINMENT BARRIER THRESHOLDS: 126 1Page Figure 9-F-43: VWR-Containment Integrity or Bypass Examples RCP Seal Cooling NOTES: Only Supplemental Purge is a filtered release and STPEGS Component Cooling Water is equivalent to Closed Cooling Water 127 1Page 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS Table H-i: Recognition Category "H" Initiating Condition Matrix UNUSUAL EVENT HU1 Confirmed SECURITY CONDITION or threat.Op. Modes: A4lALL ALERT SITE AREA EMERGENCY GENERAL EMERGENCY HAI HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.Op. Afodes: A4/ALL HS1 HOSTILE ACTION within the PROTECTED AREA.Op. Modes: A//ALL HG1 HOSTILE ACTION resulting in loss of physical control of the facility.Op. Modes: A4/ALL HU2 Seismic event greater than OBE levels.Op. Modes: A//ALL HU3 Hazardous event.Op. Modes: A44ALL HU4 FIRE potentially degrading the level of safety of the plant.Op. Modes: Note: (Q SeeSA9or CA6 for escalation of these events HA5 Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.Op. Mllodes: A-1ALL HA6 Control Room evacuation resulting in transfer of plant control to alternate locations. Op.Modes: ,4/4ALL HA7 Other conditions exist which in the judgment of the Emergency Director warrant declaration of an AleAALERT. Op. Alodes: A-tALL HU7 Other conditions exist which in the j udgment of the Emergency Director warrant declaration of an (N.O)UE UNUSUAL EVENT.Op. AModes: A4/ALL HS6 Inability to control a key safety function from outside the Control Room.Op. Modes: A4-ALL HS7 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency g4TF=SITE AREA EMERGENCY. Op. Modes: A//ALL HG7 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Geaneral Em~ergency G ERNAE GENERAL EMERGENCY. Op. Modes: A/.ALL 1281 Page HU1 ECL: of U....s..al UNUSUAL EVENT Initiating Condition: Confirmed SECURITY CONDITION or threat.Operating Mode Applicability: AI-ALL S Emergency Action Levels: (1 or 2 or 3)( L) A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by -spe4fie security shif4 supe-'visi'n) ANY of the following personnel in Table HI: Table HI: Security Supervision

• Security Force Supervisor
• Acting Security Manager* Security Manager (--(2QLNotification of a credible sec.rit. threat. CREDIBLE SECURITY THREAT directed at the site.H2(-2 LA validated notification from the NRC providing information of an aircraft threat.Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. See*4itv e'e ts-SECURITY EVENTS which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.. Security events SECURITY EVENTS assessed as HOSTILE ACTIONS are classifiable under ICs HAl., HSI and HGI.Timely and accurate communications between Security SkiP4-Force Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and OROs.Security plans and terminology are.based on the guidance provided by NEI 03-12, Templatef/r the Securiv., Plan, Training and Qualification Plan, Safeguards Contingency Plan [and -FuIiel S,-qy;-e In[s .tcd/loý,'fa JNDEPENDL)NT SPEANT FUEL STORtIGE IAUT.4LLA TION Securit, Program].EAL #1- references (4ite specific secrit, ....t s-per-ision)

Security Force Supervisor because these are the individuals trained to confirm that a see*:ity event SECURITY EVENT is occurring or has occurred.Training on security event SECURITY EVENT confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39039 information. EAL #2- addresses the receipt of a credible security threatCREDIBLE SECURITY THREAT. The credibility of the threat is assessed in accordance with (site proEccdlre) OOSDPOI-ZS-001 1.Implementing Procedure For Safeguards Contingency Events.129 1 P aEo e EAL #3- addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with (site specific procedure) OPOP04-ZO-SEC4, Guideline For Airborne (Aircraft) Threat. and Security Force Instruction SI 2700, Security Response to Airborne Threat.Emergency plans and implementing procedures are public documents; therefore, EALs s+&44do not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be is contained in noin public documents such as the Security Plan.Escalation of the emergency classification level EMERGENCY CLASSIFICATION LEVEL would be via IC HAl.HUI: EAL-1 Selection Basis: For EAL.-1, thc position of Security Force Supervisor was included since it is a 24-hour position.Normially. the event would not be reportec by the Acting Security Manager or Security Manager because the Acting Security Manager position is not normally activated until after an UNUSUAL EVENT has been declared, and the Security .Manager position is not normally activated until after an ALERT has been declared. However, reporting by the Acting Security Manager or Security Manager was included in the event these positions are staffed under unusual circumstances. REFERENCEs:

1. OERPOI-ZV-SH03, Rev. 12. Acting Security Manager OER.P0 I -ZV-TS08.

Rev. 16. Security Manager 3. 0POP04-ZO-SEC4. Rev. 10. Guideline For Airborne (Aircraft) Threat (SUNSI)4. 0SDP0 1 -ZS-001 I, .mplementing Procedure For Safeguards Contingency Events (Safeguards)

5. Security Force Instruction SI 2700, Security Response to Airborne Threat (SUNSI)

The (Site Specific Security Shlift superYis5ion) is thle title of tile OR Shift. individual responsible for The (site Speeific prOCedur~e) is the procedur-e(s) uised by Controel Roefm and/lor Security perSENnfel to deter-mine if a secur~ity thrleat is cr-edible., anid to validate receipt ofghaircaft thireat inAfrmation. Eme*rgen..y Plans and inp .leme..ting pr ..e..t.. a. public documents; the-efo4re, EALs shoulHId not 4eorort securit~; sens~itive in1formation. This iniekudes inforimation thiat may be advanitageous to a Potential adversary", suhas the partIiculars, concerning a specific thr-eat at- threat locationi. Securiity sensitive ifrainshould be contained in lion puiblic documients suchi as the Sceuriity. Pla...With due consideration given to the above developer note, EA.s may contain alpha or n-mbered e-fcren-es to selected events des.--ibed in the Seceuity Plan and associated imp flip! elemeting rSe.worded as "Secureity event 42, 45 or #9 is reported by tie (Site specific security Shift supervisionA" ECLi Assignment Attr-ibutes: 34.lA.A 130 1Pa e HU2 ECL: Nat-fieatitn of U..us.al E9.N UNUSUAL EVENT Initiating Condition: Seismic event greater than OBE levels.Operating Mode Applicability: AIIALL Eamnpe-Emergency Action Levels: (1) a. EITHER of the following conditions exist: I. Sei ..i .event gr.eate than Oper.ating Basis Earthquake (OBE) as indicated by: (site speeif.e OE-4mn'ASE1SMIC EVENT" alarm in U nit I Control Room (L-ampbox 9M01, Window E-8)OR 2. Control Room personnel feel an actual or potential seismic event.AND b. The occurrence of a seismic event is confirmed in manner deemed appropriate by the Shift Manager or Emer.,ency Director.Basis: This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.Although the "SEISMIC EVENT" alarm (0.02 g) in EAL. L.a is set below an O.B.E earthquake (0.05 ,), it does provide an indication that a seismic event has occurred. In order to determine whether an O.B.E.earthquake occurred additional indications may be needed, Determination per 0POP04-SY-00.1., Seismic Event is not practical if it takes longer than 15 minutes to perfoar.Indications described in the EAL should be limited to those that are immediately available to Control Room personnel and which can be readily assessed. Indications available outside the Control Room and/or which require lengthy times to assess (e.g., processing of scratch plates or recorded data) should not be used. The goal is to specify indications that can be assessed within 15-minutes of the actual or suspected seismic event.The EAL 1 .b- statement is included to ensure that a declaration does not result firom felt vibrations caused by a non-seismic source (e.g., a dropped heavy load). Event verificatian with external sou-rces souald not be fneeessary d uing or- follwinlg ani OBE. E~arthquakes of this miagnituide shouild be r-eadily felt by. oni site pers.nnel and recognized as a seismi. event (eg.typial lateral accelerations .inw ex.ess 4f0.OSg). The 1311 Pa e Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action mnust not preclude a timely emergency declaration. It is recognized that this alternate EAL wording may cause a site to declare an UNUSUAL EVENT while another site, similarly affected but With readily assessable OBE indications in the Control Room. may not.Depending upon the plant mode at the time of the event, escalation of the emergency classifieation level EMERGENCY CLASSIFICATION LEVEL would be via IC CA6 or SA9.HU2: EAL-1 Selection Basis: STP does not have a readily available indication in the Control Room for determining if the site has experienced an OBE. The Seismic Event Alarm setpoint is 0.02g in the vertical or horizontal position and the station design basis value for an OBE is 0.05oý. Since the Seismic Event alarm is set akt less than half of the OBE value, it cannot be used as the sole threshold value for determinin. whether or not STP has experienced an OBE.STP has implemented the alternative EAL described in NEI 99-01 Developer Notes in conjunction with using the installed indication. EAI,-. b. allows the Shift Manacer or Emergency Director to determine if a seismic event has taken place. taking into consideration the Seismic Event alarm, Control Room personnel feeling an actual or potential seismic event and other indications deemed appropriate.

REFERENCES:

I. OPOP04-SY-000 I. Rev. 8, Seismic Event 2. NEI 99-01. Rev. 6. Development of Emergency Action Levels for Non-Passive Reactors.This "site specific indication that a seismic event met o" exceeded OBE limits" should be based an the in.dicatiOnS, AlrMs Mnd displas Of Site.sp.c.ific seism.. i. mnO.I48!.... equipment. IndicatioHN described iii the EAL. shoulId elitd tO thosie that allimditl akvailable to Conltrol Ro00M per-sonnel and which can be ix'adily assessed. indications available outside the Control Room and/orWlich require lengthy times--to-ssss , r d a -suld REAt be use.d. The goal is to specify indications that Can be assessed within 15 minutes of the actual or spcedseismic event.For site-s that do not have readily assessable OBE inldicatiOnS within the Control0 RoomI, de-Velopers shou~ld use thie folloin;ig alternate [AL, (orf simiilar-weirdiig,.

a. Ceontrl Room per.sof.nel feel ani actual or potential seismic event.b. The occurrence of a seismic- event. is; cAnfir-ed in maniner deemed app:'opriate by the Shift Manager or E..mer ..........

i-.. , r-;e,_+i .. -The [A I hstatement is included, to ..ensu.e that a declaration does not.. r.eslIt Mfty " by a non seismic source (e.g,.. a dropped heavy. lead). The Shift Maniager-or Emergxency Dir-ector ma seek e-ternal verification if deemed appropiate (.g.," a call to the USGS, check internet news suces, etc.); ho...ever, the v..eriFcation action must nat pr-eclude a tim..ely e,. genev' declaration. it is :ecognized that this alter.nate [AL.= w.ording may cause a site to.declare an u ............ e.t .wrhile ano.er. site, similarly aftereted but A ith Feadik assessable "PIE indiecttiwis iii the C.entfol Room fflay not.I 132 1 P a g e Nwheli a secismie Faanitaringqsytemcm ECL Assil-nrent Attributes: -3. 1. lA apahle of deteting an OBk is wt of geF,,iee fer inainteiiaiie Of 133 IP a e HU3 ECL: Natiticati.n of Unus..al Event UNUSUAL EVENT Initiating Condition: Hazardous event.Operating Mode Applicability: A-I4ALL Exanpie Emergency Action Levels: (1 or2 or 3 or 4 o--5or 5)Note: EAL #4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. (1) A tornado strike within the PROTECTED AREA.(2) Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode.(-4:)(3jMovement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).H-L4)_A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.(3) (Site spe.if. , list of .atu.al or " hazard Predicted or actual breach of Main Cooling Reservoir retaining dike along North Wall (4sL)_Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.EAL #1- addresses a tornado striking (touching down) within the ProtectedAreaPROTECTED AREA.EAL #2- addresses flooding of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.EAL #3- addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.EAL #4- addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river water releases, dam failure, etc.. or an on-site train derailment blocking the access road. This EAL is not intended apply to 134 1 P a 2 e routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.EAL #5 addiresses (site specific des'e4,iEt*4. EAL#5- the Main Cooling Reservoir breach alone the north wall which was included because it is a credible hazard and analyzed in the STPEGS UFSAR..Escalation of the emergency classification level EMERGENCY CLASSIFICATION LEVEL would be based on ICs in Recognition Categories AR, F, S or C.iHU3: EAL-i, EAL-2, EAL-3, EAL-4 Selection Basis: N/A

REFERENCE:

I. STPEGS ULI'SAR. Section 3.4.1, Flood Protection Developer Notes:.he "Site specitic list of natu'al or technological hazard events" sheuld E!in.de other events that mia- be a precursor to a moe.. significant eve..t or cndition, and that are appri.. priate to the site .ation a Notwithstanding fl"..thle ce,,ts spe.ifie.. i..cluded as EAks above, a "Site sp.cifie list 'f natural or technologic-al hazar-d evenits" need not inckludesor lived events for wh.4ich the exNtent of the damage and the .es.lti.n c-ne ...... es can. be deter'minied w.ilhin a r "elatively short ti.. e frame. in these cases. a damage assessment can be performied soen aftei- the event, and the plant stafg will be able to identik, petential or a"tual impacts to planlt s..s.tems. anld structureS. This will enable p.omp.t n and impleffmentation of coa pewsator' or corrective measures w.ith no appreiable increase in risk to the piblic.To 4he exEtent that a shomi lived event doees cauise immiiediate and signiificant damiag~e 'o plant systenis anld structL res. it Nvill be classifiable und164r the Reogntion t .. -C ..lesser imiipact would be ex.pec.ed to cause ..l' sm.-all and loc-alized dam..age. The c.ns.equlen..es fom these;;es of e.vent. a... a "qatel asses-sed anda-ddreI ss in acc. dan.ce with Te.hnical S*. eeifica.ins. hi addiiti-en, the--eeeurrenee or- effects of the e~venit mw,- be CP ECL Assignment4,÷ "1 1 1 A ... I 2 1 1 1-')6X3mr- :ý- 0 Ing 1351 Page HU4 ECL: -f ... us4T al E.. eit .UNUSUAL EVENT Initiating Condition: FIRE potentially degrading the level of safety of the plant.Operating Mode Applicability: A-IALL Exta-mple Emergency Action Levels: (1 or 2 or 3 or 4)Note: The Emergency Director should declare the Unuis-ial Eveit-UNUSUAL EVENT promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.(1) a. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications:

• Report firom the field (i.e., visual observation)
• Receipt of multiple (more than 1) fire alarms or indications
• Field verification of a single fire alarm AND b. The FIRE is located within ANY of the 4b4&-wing plant rooms or areas in Table 1-144:ýSitP 1;PRP-if41-lir# AfRIA14t r-A-1111 --1-N Table HA4: Plant Roorns/Areas o Mechanicali/Electrical AUxil iarv Building (MEAB)* Fuel I landling Building (FI1B)" Reactor Containment Building (RCB)* Essential Cooling Water Intake Structure (ECWIS)o Isolation Valve Cubicle (IVC)* Diesel Generator Building (DGB)wit-hye a'4 Turbine Genierator Buiklding Fuel Handling Building P. "ne.t:- C":'ii.n h::n.rn R:i!,1lino R-ildilw.........E&Sential Cool ing Water Intake Str'cture Isolation Valve Cuibicle Diesel Genlerator Building CirculIAtigater-Inake Str-ucture (2) a. Receipt of a single fire alarm (i.e., no other indications of a FIRE).AND b. The FIRE is located within ANY of the 41oPving plant rooms or areas in Table 1I43: 1361Pag-e (site :;pecife list' of Plant r'ooms or- ar-eaq)witehlyard Tur-bine Gjener-ator Building Meehanieal,lEleetrieal

~A~teiliai-y Building Fuel ,andlin ,uIl--ing Reaetor: Containment Buildiriig Essential Cooling Water- Intake Structure Isolationi Valve Cubiele Diesel Generator Building Circulatinig Water- hitake Struttur~e AND c. The existence of a FIRE is not verified within 30-minutes of alarm receipt.(3) A FIRE within the lSFSI OR pIlant [it-: ' .,,hn . ......... .. , .]PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication. (4) A FIRE within the ISFSI OR plant [for planis with an. .SFS. ou.side the plant Pr.tected rea]PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.EAL #1 The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.EAL #2 This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarn. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed. 1371 Pa g e A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.If an actual FIRE is verified by a report friom the field, then EAL #1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. EAL #3 In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant or ISFSI PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. 44i6 basis extends to a FARE- ocellrr4in wihntePOETD-L' fa SS oae udeth an PROTECTED rRg. nt with. a POES! ousd r e&,-, ....... .,. ., the plant Pv.teetedrea] EAL #4 If a FIRE within the plant or ISFSI_ for*- .. h a. .,, 4 ,the t.., ,"he lant P t Area]PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery RECOVERY or investigation actions.Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part: Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and EXPLOSIONS." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of l-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train 1381 P a e (G.2.c). As used in EAL #2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.Depending upon the plant mode at the time of the event, escalation of the emergency classifieation level EMERGENCY CLASSIFICATION -weu4dLEVEL would be via IC CA6 or SA9.HU4: EAL-I.b, EAL-2.b Selection Basis: The plant areas or roolms listed contain SAFETY SYSTEM equipment.

REFERENCES:

I. OPGPO3-ZF-O00 I. Rev. 26. Fire Protection Program 2. STPEGS UFSAR. Rev. 16,_Section

7.4. Systems

Required for Safe Shutdown 139 1 P a g e The '"ite speei fie lit 4f planit r-ooms S YSTEM equipmeiit. As nioied in the FAIL.- anid Basis see!Prote~'edkie -:1-1-1A A ;. 1 .- , .2 1 i oi. areas" Should spe.it.. those rooms at- areas that contain SAFETY ion,. include thle term IsfEsl if the site has an ISFSI oultqide thle Plant ttF HtE!S: ...A I 140 e HU7 ECL: Notification of Unusual Event UNUSUAL EVENT Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a (NOýUE.Operating Mode Applicability: AI4ALL EAitniF!e-Emnergency Action Levels: (1) Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to FAC1LITYfeaci-4-y protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the.lassifieatio n ll ..EMERGENCY CLASSIFICATION LEVEL description for an NOUEUE.HU7: EAL-1 Selection Basis: N/A

REFERENCES:

N/A 141 1Pa e HA1 ECL: A4eq4 ALERT Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.Operating Mode Applicability: AI1ALL Example-Emergency Action Levels: (I or 2)(1) -....A HOSTIL6E ACTION is eeecurringe or has ecculfed within the OWNER CONTROLLED AREA as repaitcd by the (siteSj,.o, spec- u i-s S v .c S-pep.'iszp OR Acting S t Mane. A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by ANY of the following personnel in Table HI: Table HI: Securill Supervision" Security Force Supervisor

• Acting Security Manager* Security Manager 14A validated notification from NRC of an aircraft attack threat within 30 minutes of the site.Basis: This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Spe-ý4 FV'o c! ... S...... Jn..t.allatioh" INDEPEAVDE.NT SPENT FL/EL STOt-4GE INSTALLA TION Security Program].As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering).

The A-4e-ALERT declaration will also heighten the awareness of Offsite Response Organizations, allowing them to be better prepared should it be necessary to consider further actions.This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small 142 1P a R e aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.EAL #1- is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This inludes a.'..a.tio. afaifSt all .S.SoI that is l..ated outi. de the plant AREA.EAL #2 addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with (site specicfi procedure) OPOP04-ZO-SEC4. Guidelines for Airborne (Aircraft) Threat, and Security Force Instruction SI 2700, Security Response to Airborne Threat.The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency.Emergency plans and implementing procedures are public documents; therefore, EALs not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information sheold be is contained in non public docu'ments such as the Security Plan.Escalation of the emergency' classification level EMERGENCY CLASSIFICATION -wau4LEVEL would be via IC HSI.HAl: EAL-1. and EAL-2 Selection Basis: The EAI..,s are taken fiom NEL 99-01, Rev. 6. For EAI.-1. the positions of Security Force Supervisor OR Acting Security Manager were included because either of these positions could be activated prior to meeting this EAL. The Security Force Supervisor is a 24-hour position and the normally the Acting Security Manager is activated after an UNUSUAL EVENT has been declared. The Security Manager is also included t4--although this position is normally activated after an ALERT.

REFERENCES:

I. OERPO I-ZV-Sf 103. Rev. 12. Acting Security Manager 2. OERP0 I-ZV-TS08. Rev. 16, SecuritN Manager 3. 0POP04-ZO-SEC4, Rev. 10. Guideline For Airborne (Aircraft) Threat (SUNSI)4. Security Force Instruction SI 2700, Security Response to Airborne Threat (SUNSI)Developer-Notes: 143 1P a g e prvi..on..f.the .shift seeiý fre-.Fmeirgency plan5 and implementingz pi-. .ai-s af-. publi. documents; therefre. EALs should not r tSeouritv sensitive information. This includes infor.mation that malp' be atdantageous to a potetia adersrysuc as the paticiulars conieerninig a speecific threat or threat locationi. Seeurit'.senlsitive infoorm-ation Should be contained in noen puiblic documents suchl -AS $heR- $Securjity Plall.With .due on.sideration given t.othe above develope.. note, EALs may .ontaif. alpha o0! numblered references to) s.ele.ted e"ent.s dS..ibed ill tihe Secuity Plan. associated imaplementing p-ocedues. Such .. sho.. ld not con.ain a r-ecognizable descr.iption of.. he event.. Foi example. an ..AL ma' be..vor.ded as "Security event 41t.. 5) or 49 is r.eported by the (site speci.. c sec.. .rity shi!", supeirvisio-)." See the related Developeri Note iii B, fei- gu..idanc. he developmen.t o fa scheme definition for the OWNER CONTROLLED AREA.144 1 P a , e HA5 ECL: le# ALERT Initiating Condition: Gaseous release impeding access to equipment necessary for normal plant I operations, cooldown or shutdown.I Operating Mode Applicability: AIIALL I ~ Exam-ple Emergency Action Levels: Note: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. (1) a. Release of a toxic, corrosive, asphyxiant or flammable gas into the Control Room or A-N4-ANY of the f.lew-ing pant rooms or areas listed in Table H3/R2: (site spehific list of plant r..ms or areas with entF- related Mode appli;abi lit, identified) Tafbiiie GeHei!ateiý Buildifiý414G134 jS0jfttiE)f1 3,701- C-h-1p F-Hel Handline Buildine (FHB)AND b. Entry into the room or area is prohibited or impeded.I TABL F W/I2: PI-rnt Ar,- P- uridrina Arpz RCB RHR Heat Exchanger Rooms 0 0 rI MAB 51 ft Room 335 EAB Roof, MCC 1G8, 4.16KV Switchgear Rooms o Ln) EAB 4.16KV Switchgear Rooms 145 1 P a g e Basis: This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant.An A-eA ALERT declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release.Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). An emergency declaration is not warranted if any of the following conditions apply.* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release)..--For example, the plant is in Mode 1 when the gaseous release occurs, and" the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.* The gas release is a planmed activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing)." The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections)." The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.This EAL does not apply to firefighting activities that automatically or manually activate a fire suppression system in an area. or- t i -ntentional ine.. tiing of .,ntainment (AI3R only).146l P a, e Escalation of the emergency classification via Recognition Category AR C or F ICs.level-EMERGENCY CLASSIFICATION LEVEL would be HA5: EAL-1 Selection Basis: The areas listed in FAI,-I apply to areas that contain equipment necessary for plant operations, cooldown, or shutdown. Assuming all plant equipment is operating as designed, Normal operations and safe shutdown equipment operation is capable from the Main Control Room (MCR). The plant is able to transition into a hot shutdown from the MCR. therefore H3/R2 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary following entry into hot shutdown (establish Residual Heat Removal shutdown cooling, disable operation of charging and ECCS equipment, and limit dilution pathways) and subsequent entry into cold shutdown (disable operation of ECCS equipment). After achieving cold shutdown it is assumed that the plant will be maintained in a cold shutdown condition.

REFERENCES:

1. OPGP03-ZF-0001, Rev. 26, Fire Protection Progrzam 2. STPEGS UFSAR. Rev. 16. Section 7.4. Systems Required for Safe Shutdown 3. OPOPO3-ZG-0008.

Rev. 56. Power Operations

4. OPOP03-ZG-0006, Rev. 54. Plant Shutdown from 100% to Hot Standby 5. OPOPO3-ZG-0007.

Rev. 71, Plant Cooldown DevelopertNot es: The "site specific list Ofplant roomas or areas with entr-y r-elated mode applicability identifled"' shouild ope.OC.edures ..sed Ie.r norm"l .plant opei.Aon, co.ldown shud w!n Do not inClude rooms or areas Mn Whlieh ac-tionls 4fa eantingent Or emergency natur'e would be perfor-med (e.g.. an actioni to addre&ss an off nor~mal or- emer-geney eondition stich as efflergne rI pirs creciietie mieasures 0ir emergeiiey. peatins. In additioll, the list shoumld specifyf thle 1)ant mo1de(s) durling which enItry would be reOquired for eac h room or area.The list should fiot inlude rooms el- areas forw.ih... entry is ..ui.ed solely to perform actions ofan administr-ative or- record Ikeepinig nature (e.g.. normal rounds Eor routine inspections,) .The list need not incluide the Control Room if adequiate engineered safýety/.designi tfatures are in place to pr-e-kmDe-a-Cntrol Room e-oacuaoie -4uo4e release ofa hazardet.sgas. ma;, inclunde bu are not limiited to. capability to draw air- froem multiple air- intakes at deifbent and separ~ate location-ine and outer- atmoespheric boundar-ies, or hiapblt to acquir-e and mainiain positive pr-essur-e w~ithin the If the equipmenit ill thle liSted room1 Or are-a wkas alre@ady in!operable, Or out of SerVice, befor-e the evenAt.....u..ed. then. no emer.,ene.. should be declar.ed since the event will ha n.o adver.se impact b..od that aready allowed by Tecfhni.al Specications at te time of the even.t..F:CL AsI enrnn; Attribbites:3.I2B 147 1 P a g e HA6 ECL: ANe4 ALERT Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations. Operating Mode Applicability: A44ALL Example Emergency Action Levels: (1) An event has resulted in plant control being transferred from the Control Room to (scite speeiPe ie -Auxiliarv Shutdown Panel Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room., will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges. Escalation of the erf, egenlcy classification levl E4MERGENCY CLASSIFICATION -wo+d4LEVEL would be via IC HS6.HA6: EAL-1. Selection Basis: The Auxiliary Shutdown Panel (ASP) is identified in OPOP04-ZO-0001. Control Room Evacuation. as the location where plant control is transferred in the event of a Control Room evacuation.

REFERENCES:

1. Procedure OPOP04-ZO-0001, Rev. 35. Control Room Evacuation Dcvelopcr Ntes: The "s:.ite pcific r-emote "hu"itd....n panels and Iccal controjl .tatiai."'

arc the panels and centtrol statioas.---rncd iin plan; precedures used to eeoldewyn and shutdeNwn the plant firom a lee-ation(s) ouitside the cofft+ReýA it .;I, , , -'2 1 ') E)i 11 , 0 148 1 P a R e HA7 ECL: A-44 ALERT Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of an AMe4ALERT. Operating Mode Applicability: A4-ALL.Nample Emergency Action Levels: (1) Other conditions exist which, in thejudgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event SECURITY EVENT that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION.A-wyANY releases are expected to be limited to small fractions of the EPA P "Ectevz Actio" (-*iadeline PROTECTIVE ACTION GU IDEL INE exposure levels.Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emcrgznc.y elazsification l.vcl EMERGENCY CLASSIFICATION deserpenLEVEL description for an Aje4-ALERT. HA7: EAL-1 Selection Basis: N/A

REFERENCE:

N/A 149 1 P age HS1 ECL: SiteUea Emer.gene.. SITE AREA EMERGENCY Initiating Condition: HOSTILE ACTION within the PROTECTED AREA.Operating Mode Applicability: AI-ALL ENample-Emergency Action Levels: (1) A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site spe.ific.se....itoy. shift supe.. ision). -nny ANY of the followine personnel in Table HI: Table HI: Security Supervision

• Security Force Supervisor o Acting Security Manager* Security Manaler Basis: This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security. Plan, Training and Qualification Plan, Safeguards Contingency Plan [and izde*qeiik- F: .,e! .....ag. 4...a..aiiw--. INDEPENDENT SPENT FUEL STORAGE INSTALL4 TION Security Progran].As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergefncy SITESITE AREA EMERGENCY declaration will mobilize ORO resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions. This IC does not apply to a 1 1-STIIE-ý ACTION dire.ted at ani PROTECTED -ARE-NA l.eated....oide thie Plant PROTECTED AREA; .u... an attack should be assessed , W. H HA i It anot app4y-4e,-incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.Emergency plans and implementing procedures are public documents; therefore, EALs sheal-do not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information s-otkld be is contained in n'ii public doclumnenis,+e4-as-the Security Plan.150 1 P ag e Escalation of the emergency classification level EMERGENCY CLASSIFICATION LEVEL would be via IC HGI.HSI: EAL-1 Selection Basis: The positions of Security Force Supervisor, Acting Security Manager, and Security Manager were included since any of these positions could be activated prior to meeting this EAL. The Security Force Squervisor is a 24-hour position, the Acting Security Manager is activated after an Unusual Event has been declared and the Security Manager is activated after an Alert is declared.

REFERENCES:

1. OERPOI-ZV-SI103, Rev. 12. Acting Security Manager 2. OERPOI-ZV-'I"SO8.

Rev. 16. Security Manager Develepcr-Notes: TPhe (Site spepcifi security "hift super-vision) is the title, of the An 4hif indiv.idual responsible for supc'.'iionof 4hc ell shift security 4ýforee Emergelcy' plans and ,impleenetng procedures are public documents; therefore, s should not incrpoateSecuritN,' sensitive infor-mationi. T!his includes, inforemationi that may, be advaiitageeus to a p.t..t.a adversary.. suH. aS ,he P articuflrS conerning a spiic th.e. or:thre" lation. Security sensitive Finformation should be coentained in non public documents SuhA as the SecuritN' Plan.With &w .......a.. : i','c--n to thle above de-Veloper n.te.. EA-s may. cntain alpha 0r nimber-4 retýferenes to selec-ted events described in the Seeurity Plani and associated impleetn procedue.Such r-et~ferenes shouild not contain a reecognizable description ofh0 vn.Freape nELmyb worded as "Security event .2. 455 or" ,9 is reported by the (site specific ect shift super vi sion).+See the rela Developer Note in Appendix R.. for" guidaimte on the development of a sceme definitio-n Por the PROTECTED AREA.ECL Ass-,nment Attrbutes:

3. 1.3.. D 151 Pag- e HS6 ECL: Site Area Emergency SITESITE AREA EMERGENCY Initiating Condition:

Inability to control a key safety function from outside the Control Room.Operating Mode Applicability: A-IALL E*:imple Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergency SITESITE AREA EMERGENCY promptly upon determining that (site specific num-ber of milutes) 15 minutes has been exceeded, or will likely be exceeded.(1) a. An event has resulted in plant control being transferred from the Control Room to (site speei-4e reI!!ote shutdon ... el. and l... cntr. l ," atio:s).the Auxiliary Shutdown Panel (ASP).0ae followinig locations: AND b. Control of ANY of the following key safety functions in Table 11.2 is not reestablished within (it .....cific number of minutes) 15 minutes in Modes 1, .or 3 ONLY.).Reaetivt"v control" ,ooling, [DWR] / RPNV wkater. level [Br PRI" RCS heat removal Table H2: Safety Functions* Reactivity control* Core cooling* RCS heat removal Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.The determination of whether or not "control" is established at the remote safe khutdoen ) n Auxiliary Shutdown Panel is based on Emergency Directorjudgment. The Emergency Director is expected to make a reasonable, informned judgment within (the site specific tim.e ... t..ansr.) 154-0 minutes whether or not the operating staff has control of key safety functions friom the remote safe shutdown location(s). Escalation of the emergency classifieation lev'el EMERGENCY CLASSIFICATION LEVEL would be via IC FGI or CG1.152 P a g e HS6: EAL-1. Selection Basis: The Auxiliary Shutdown Panel (ASP) is identified in OPOP04-ZO-0001, Control Room Evacuation. as the location where plant control is transferred in the event of a Control Room evacuation. The 15 minute timeframe to control the key safety functions is identified as site specific information. The mode applicability conditioning statement for Table H2 is based on the Technical Specification Operability requirement for the following functions of the Remote Shutdown System: " Core reactivity control (initial and lonv term)" RCS pressure control* Decay heat removal via the AFW System and the SG safety valves or SG PORVs" RCS inventory control via charging flow. and" Safety support systems for the above functions.

REFERENCE:

-Procedure OPOP04-ZO-0001. Rev. 35, Control Room Evacuation I.2. Technical Specification 3.3.3.5 Remote Shutdown System 153 1 P aiz e

3. Developer-Notcs: 4. The "site specific remote shuttdown paniels and loeal eontroil stationis" arc the panelsan contrtol stationis refei-cnccd in plant proeedur~es uised to cooldown mnd shutdown the planit froem a locationi(s) outside the Control Room.5. The "site spec-ific number of minutes" is the tome in which plant control must be (or i ecipeeted to be) reestablishied at ani alternate location as descr-ibed in the site specifc &-re response anialyses.

Absent a basis in the site speeifie analyses, 15 minuites should be used-.Anoether time period may be used with approepriate basis/justification-.

6. EGLm Assignment Atifributes:.

21. 3.B 154 1P a g e HS7 ECL: Site Area Emergency SITESITE AREA EMERGENCY Initiating Condition:

Other conditions exist which in the j udgment of the Emergency Director warrant declaration of a Site, Aea fEmegene' SITESITE AREA EMERGENCY. Operating Mode Applicability: A-HALL F*a-vnpk-Emergency Action Levels:_Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional darnage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. AnyANY releases are not expected to result in exposure levels which exceed EPA eeiP-veAetion G* PROTECTIVE ACTION GUIDELINE exposure levels beyond the .44e nd.ai+y-SITE BOUNDARY.Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergen.y classification level EMERGENCY CLASSIFICATION LEVEL description for a Site Area]Fieige+iy SITE AREA EMERGENCY. 14S7: gAL-I Selection Basis: N/A

REFERENCE:

N/A 1551 P a e HG1 ECL: General Emergeney GENEPLGENERAL EMERGENCY Initiating Condition: HOSTILE ACTION resulting in loss of physical control of the F FACI ITY.Operating Mode Applicability: A-4ALL E1nniple-Ernergency Action Levels: (1) a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site spccifi s.... supe:viwisn). aIn ANY of the following in Table I11: Table HI: Security Supervision

• Security Force Supervisor
• Acting SecuritV Manager o Security Manager AND b. EITHER of the following has occurred: I_. ANY of the following safety functions in Table H12 cannot be controlled or maintained in MODES 1, 2 or 3 ONLY.Reaetivity efntro~l CorFe eeelin~[Pgl, RP% water- lev~el [XVR]RGS heat r-emoval Table 1-12: Safetv Functions-Reactivity control* Core cooling o RCS heat removal OR-k2.Damage to spent fuel has occurred or is IMMINENT.1561 P a g e Basis: This IC addresses an event in which a HOSTILE FORCE has taken physical control of the 4weity FACILITY to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions.

It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEt 03-12, Template for the Securio, Plan, Training and Qualification Plan, Safetguards Contingency Plan [and I~tkp:nden&Sptw* Fwel Storg ,i,o.. ...fIN.-DEPENDENTS PEINT FU, EL STORA4GE INS TAILLA TION Sec uni., Progr'am]. Emergency plans and implementing procedures are public documents; therefore., EALs shuld__do not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information sh'Eomd is be is contained in non public documents such as the Security Plan.HGI: EAL-1 Selection Basis: The positions of Security Force Supervisor. Acting Security Manager, and Security Manager were also included since any of these positions could be activated prior to meeting this EAL. The mode applicability conditioning statement for Table H2 is based on the Technical Specification Operability requirement for the following Functions of the Remote Shutdown System: " Core reactivity control (initial and long term)" RCS pressure control* Decay heat removal via the AFW System and the SG safety valves or SG PORVs* RCS inventory control via charging flow, and* Safety support systems for the above Functions.

REFERENCES:

I. OERP01-ZV-SH03. Rev. 12. Acting Security Manaer 2. 0ERP0I-ZV-TS08. Rev. 16, Security Managzer 3. Technical Specification 3.3.3.5 Remote Shutdown System Devekopei-Notes: 1571 P a e The (site spe.i..c seemrity shi.. i the,4itle of.the an shift ... , , 1uperi0io Of thle OR Shift Seurit', fOrce, Emiiergen plae n ..d implee;tig pr'ocedures are public document.s; theriefoire. EALs should n MnOrFporate Security sensitive informIfationl. ThiS iclu~kdeS infor-mation thatI may be advantageous to a potential adversar-y, suchi as the particuilar-s concerining, a specifi thr-eat ot- threat loeatien. Securityýsensitive infor-mation should be containled in ne!n Public documents Such as !he sccuity45 Plan., With dute consider-ation given to thie above developer note, EALs may contain alpha or numibere reAef-erfenes to selected cvcnP-ts delscribed in the Securfity Plan and associated imp!ementing.procedures. Suchi i.:frences should not cotain a recognizable description ofthe event. Forh exPmRple, an A.L may be wor-ded as "-Secur~ity event 42.. #-5 or 49 is i-epor-ted by the (site :;pecitic security shift supervision-)." definition for- the PROTECTED AREA-.EC-A-Y 158 1 P a g e HG7 ECL: General Em'crgency GENERAL EMERGENCY Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General EmfergnyG ENERAL EMERGENCY. Operating Mode Applicability: A41ALL Emergency Action Levels: (1) Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the fFaeilitNTACILITY. Releases can be reasonably expected to exceed EPA Protective Acttion C'ideline PROTECTIVE ACTION GUIDELINE exposure levels offsite for more than the immediate site area.Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency' classification level EMERGENCY CLASSIFICATION LEVEL description for a GenefaI E.e.ge.ey GENERAL EMERGENCY. HG7: EAL-I Selection Basis: N/A

REFERENCE:

N/A 1591 P a g e 11 SYSTEM MALFUNCTION ICS/EALS Table S-I: Recognition Category "S" Initiating Condition Matrix UNUSUAL EVENT ALERT SITE AREA EMERGENCY SU1 Loss of aI4ALL offsite AC power capability to emergency buses for 15 minutes or longer.Op. Modes:.OSt,;atdiy

T, 1,2.3,4 SU2 UNPLANNED loss of Control Room indications for 15 minutes or longer.Op. Modes

Pii&ei-.1,2.3,4 SU3 Reactor coolant activity greater than Technical Specification allowable limits.Op. Modes: Po.ei, fluc'ti;:, 1.2,3,'4 SU4 RCS leakage for 15 minutes or longer.Op. Modes: P#owe-('.S ..an lb , fiT.- 'l, l .......ow L ,2.3.4 SU5 Automatic or manual (trip -[4fails to shutdown the reactor.Op. Modes: P.'wei 6ýeoe ...... 2 .. 4 SA1 Loss of aI4ALL but one AC power source to emergency buses for 15 minutes or longer.Op. Modes: Po;w'i" SJ,.,ar:, up .-, N , IL.,: SA2 UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.Op. Alodes: P,.e", Ope;'wi&HT Sttup.... -ot Stn..y., H4 , Shwý&tki1,2,,j4 SS1 Loss of a4ALL offsite and aM4ALL onsite AC power to emergency buses for 15 minutes or longer.Op. Modes: P.we" QO .....ioo., .......lq, 4Wt 1.2,3,4 GENERAL EMERGENCY SG1 Prolonged loss of a4ALL offsite and alALL onsite AC power to emergency buses.Op. Modes: Powe+'- St;..t.p,.

Hl , ,S ..b, H. : Shmtk"wt-L2_3 4 SA5 Automatic or manual (trip PWR]" / r rDal [BW.R") fails to shutdown the reactor, and subsequent manual actions taken at the reactor control ee-sa1L-s-anels are not successful in shutting down the reactor.Op. Modes: Powero- en 1,2 SS5 Inability to shutdown the reactor causing a challenge to (core cooling [#R]RPV Nrater level [+BR])or RCS heat removal.Op. Modes.: P 0 wei 160 1 P aa e 161 Page 162 1 P a g e Table S-i: Recognition Category "S" Initiating Condition Matrix (cont.)UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY SU6 Loss of a-4ALL onsite or offsite communications capabilities. Op. Modes: P&wei-Pe ... ndk, HaiG sh.fdifw,. ...L2,34 SU7 Failure to isolate containment or loss of containment pressure control. [42414R]Op. A-r.d.. Po.i. .ot.l..aeiou, saduif,1.. ....1,2.3.4 SS8 Loss of al-4ALL Vital DC power for 15 minutes or longer.Op. Modes: P.we;O'c1,2, ',. 5;tz;p, Hot Sc ...... M,,c, Sh:w:duwn SG8 Loss of a-4ALL AC and Vital DC power sources for 15 minutes or longer. Op. Modes: Powe*Op12otia.4 , S:or'up, Hct 12,_,,4 SA9 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.Op. Modes: Power 1,23oio,4 5u p o 1631 P a e Sul ECL: Notification of Unutsual Ev eit IUNUSUAL EVENT Initiating Condition: Loss of *IALL offsite AC power capability to emergency buses for 15 minutes or longer.Operating Mode Applicability: Po-ewr 0 Oeffl tion,, startup. met Standby. t S......O\\ nl 1.2, 3,4 Example Emergency Action Levels: Note: The Emergency Director should declare the Unusu-al Event, UNUSUAL EVENT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Loss of A-bbALL offsite AC power capability to (site pe three 4160V AC ESF Busesgasses for 15 minutes or longer.-Mle emet-gency buses).. aIALL Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency buses. This condition represents a potential reduction in the level of safety of the plant.For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.Escalation of the via IC SA1.eme..gency ,lassi4iatoin leve .EMERGENCY CLASSIFICATION LEVEL would be SUI: EAL-I Selection Basis: N/A

REFERENCES:

1. OPOP04-AE-0001, Rev. 44. First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus 2. 0POP04-A.E-0004, Rev. 15. Loss of Power to One or More 4.16 KV ESF Bus 3. OPSP03-EA-0002.

Rev. 32, ESF Power Availability

4. Drawing OOOOOEOAAAA, Rev. 24. Single Line Diagram, Main One Line Diagzram.

Unit No. I&2 Developer- The "site specifie emer-gency are te buses fed by off-ite or emergency AC po'el" s ources that1 supply.' pwer to the elcet.;,ical dibstri.bu.,,ti.n system that powers SAFETY SYSTEvS. There is typi"ally .efnergeiicy bus per tr-ain of SAFETY4 SYSTEMS'.4-At mualti un~it stations, the EALs may rdt opnstr measur~es that are pi-ocedural iied and can be implemiented within 15 miues onsider capabilitieS suceh HS powerf soHIe ui- ýi~ tie, 'si 164 1 P a g e generfatOFS. OtherF poVer SOUrceAS deScribed O in aoral Fr efergenc, beran poedues et.Plantis that have a poeuazed apability to suipply, offsite AG pEower to aii affected uniit via a el'oss tie to a cmpaMIAn fl ay' cr-edit thiS p)OWfr SE*ure in the EAL provided that the planned cr-oss tie strategy EC!I..Assilmellt p," e'ts of 10 CFR 0 A41.1hrRrI -1 1~' 1 A.165 1 P a 14 e SU2 ECL: Notiifiation of U.:2H ql :.'2-Event UNUSUAL EVENT Initiating Condition; UNPLANNED loss of Control Room indications for 15 minutes or longer.I Operating Mode Applicability: Power ()pet.ation.. Startup. Hot Standby. Hot Shutdow-i-'"4 E**mpke-Emergency Action Levels: Note: The Emergency Director should declare the UNUSUAL EVENTUn.sua .Event promptly upon detennining that 15 minutes has been exceeded, or will likely be exceeded.(1) An UNPLANNED event results in the inability to monitor one or more of the following parameters in Table S I from within the Control Room for 15 minutes or longer.[BWR paramieter list] [PWAR paramieter list]Reactor Power Reactor Power RPV Water Level RCS Level RPV Pressure RGS P .......Primary Containment Pressu.r.e in C)ore/GCre Exit Tem.peratur.e Supprpesi.n Pool Levl Levels in at least (site specifie number) twe steam generators Suppression Poel Temperatur-e Steamn Generator Auixiliary or Emefgeney Feed Water Flow Table SI: Plant Parameters" Reactor Power* RCS Level" RCS Pressure" Core Exit Temperature o Levels in at least two steam generators o Steam Generator Auxiliary Feed Water Flow Basis: This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a*more significant event and represents a potential degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the 166 1P a 2e Control Room sources for the given parameter(s). For example., the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling [PTJ'] / RPV level [B:jR] and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for r'eactor vessel .e'eIRCS level [PDWR12" / D\P/V waterrrWPr1cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classifi.cation le-el EMERGENCY CLASSIFICATION LEVEL would be via IC SA2.SIJ2: FAL-1 Selection Basis: The parameters listed were from NEt 99-01. Rev. 6 with the exception of steam generators. Two steam venerators is a site-specific parameter for the minimum number of steam generators needed for plant cooldown and shutdown.

REFERENCES:

1. OPOP05-EO-E020.

Rev. 11. Faulted Steam Generator Isolation 2. OPOPO'-EO-FRHl _ Rev. 23. Response to Loss of Secondary Heat Sink Developer-Notes: in tile PAIR paramleter-list columln, the "site specifi nainber'" shauld r-eflect the mininium number ct steami generaters necessar~y bfr plant eooldown and shutdown. This criter-ionl may, ak'is specifyý whether tile leNvel valuie should be wkide r-ange. niarro-w ranige or lboth. dpnigUpon the moneitoring -r11!e~]inemets ill emer~gency oper'ating proce~dmres. Developers may, specify' either pressurizer' er reactor- vessel level in the PW'R parameter' columnn eniti.' Afr R-CG-S-L-e-ea. The num.ber, t),pe, location. and la...ut of.Control ROomf. ind.icaion.S. and the ran.ge of possible failure modes,. can hallenge the ability oF an ope..ato ' to accurately' determine, w;ithin the time per.iod m'ai"ab freergeny classificationi assessments, if a specfitc percenftage Of inidicatiOns-haý' beenI je.t. The approPach used inl this E .facilitates pr~ompt and accur-ate emei'geney classification assessments by'feetisiing on the insdicationis fori a sele Eiea. SuOSef OF PRI'aramel'erS. 1671 Pa g e uhgon the vilbi o4h-p~eti1e-1para.. .. vaue, " "t ad ')e&.the 1"A reeceizes ..d ac..mm.d.t.. the wide va.. .; of indications ii ntuclear power plant -.... i-e iniiulmeter-valuie ()r computer, gr~oup display, etc-.E4 lp! oalint annunciators, v.11 !be ev~aluiated fbir reportability ini accordEanc~e With 10 GFR 50.72 (and thie associated ...ida.e in NUREG 02-2), and rep.rted ifit .igni.le. ly impairs the capabili, y to Pefjormi-emeiroency assessments. Comipensator:, measuires for a losof annunciationi can be rea~di ly knplemented and may i"c"d incr.a.ed monitorinlg Of...ainl co...tr.ol boar..s and mo.e fr.equent plant r.ounds by non.HCelce.d J.pejattoo;. Their' fflcrtia-funtion notwitoadi, annunciators do not provNidle the parameter" values o0" specifici .o ..ponen.t status intbrm..ation used to apei.a.e the plant., or proc ess th,,outgh AOPs o-r EOP-4. Based on these considefations. a loss of annunciation is conSidfred to be adequiately addr-essed by r~eportabijity cr-iteria, and therfefore not incluided in this IC and EAL, WWI !respect.........it eity, the r.espon.se to a lass of r .adiation onitor.ing. data .e., p..e.... or effluent monitor.values) is consider.ed to be adequately bounded b. the requ.ire.ents o 10 C-FR 5-40.72 (anld associated guiidance in N LREG I 02-2)N. The roting of this event will ensure adequa-te plant staff and NRC an..',.d the establishm.ent of,.. approprFte " ompensa.or. aseand corrective actions. in addition, a loss Hilolitorlndllata, by itself: isO Ill.ot aI pecurlsl tle a 'smmiatevent. Personnel at sites that have a Faiklure Modes and Effýects Analysis (4449A.) inclAuded Withini the deSignl basis efa digital !&C- systemi should conisider, the E-MEA informfatiOn Wilhcn developing thieir site specific Dule to cllanges, in tile colnfigurations SAFTY SVSTEMS, inc;luding assocated instrum..entat.ion. and;ndications, duringthe cold ..down.., and o0 analogouS K; is -Fo these moedes of operation+. ECL6 Assignment Attib,'-utesý 3.1. 1A 168 1P a e SU3 ECL: Notifieation of Unusual Evet UNUSUAL EVENT Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits.Operating Mode Applicability: PoN.. Opeati n, St Action Levels: (1 or 2)urtup. Hot Standby, Het Shbtdevvin 1,2,3.4 (1 (St peii va4e.racilatioii ..i......r4RT-8039 readingfeadings greater than 30 uCiicm 3 (s4ie/Ceelfl-e (2) Sample analysis indicates that a reactor coolant activity value is greater than an allowable limit specified in Technical Specifications.

• Greater than I LCi1-4 4/Pgm Dose Equivalent I- 131* Greater than 100/ EE bar LtCi /gi gross activity Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications.

This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.Escalation of the emer.gency elas;ifieation level EMEF via ICs FA 1 or the Recognition Category AR ICs.,GENCY CLASSIFICATION LEVEL would be SU3: EAL-1 Selection Basis: RT-8039 is the Failed Fuel radiation monitor and samples via the CVCS letdown line. [hle value 30 LiCi/cm 3 is the reading, that is equivalent to I LiCi/gm Dose Equivalent 1-131. The monitor value in this EAL is the calculated monitor response if the RCS activity were equivalent to I LCi/gln Dose Equivalent 1-131. The value is based on Calculation STPNOC013-CALC-003. The value used in this EAL was conservatively truncated by approximately 5% to ensure the value is readily assessable. SU3: EAL-2 Selection Basis: The Technical Specification limits for RCS activity is greater than I pCi/gnm Dose Equivalent 1-131 or greater than I 00/E bar L[Ci /ýLm gr'OSS activitY.

REFERENCES:

I. Calculation No. -STPNOC013-CALC-003 Rev. 1, Gross Failed Fuel Monitor Response to Rise RCS Activity (RT-8039 EAL Threshold)XXX

2. STP Technical Snecification Section 3/4.4.8 Snecific Activity.I 1 ........ ...1691 P a e For EAld 1 f[iner tihe radiat ies n mon- i to r(s) tha- ay he utsed to readi rly ideinti f- wh"en RTS acthiv levels exceed Technical Speci'iattion allowable limits. This EAL may be deeloped using diffelxfnt m.eh.d .An4d sites q 'hould use ei.tin.g .apabilitieS tO addfeSS 4t (e.g., deVelopMen.

.Of nv. ..apab.iliteS iS not required). Extamples of existing methods...ap.abilities ifude: " An inistalled r-adiation moanitor ein the letdown system or aiv eje fa implenientable conver'sion calculation capability. The mcniater r-eading values should corrFespond to an RCS activity level approxdimately at Techinil SpeiiCatio-nv allowable limits.I* the is no existingmetPhd/,apabilit' afo deterMining this EAL, then it shoAld not be included. .ealuatieii will be based on EAk #2.For- EAL42 Developers, may rewor-d thle EAL to inluC"de the reactOr coolant activity paaetrs specified ini Techniical Specifieaiitios anid tie associated allowable limit(s) (e.g, values 1for do~se equiivalent 1 13 1 a1d gross, activity, time dependent of iranisient values. etc.). it' this approeach is selected, all RCS activity allow'able limits should be included.6 ECL Assittnmient Aitiribuites: -3.1. k 1 A and 3. 1. 6B 170 Pa, v e SU4 ECL: Notification of Unumual Even't UNUSUAL EVENT Initiating Condition: RCS leakage for 15 minutes or longer.Operating Mode Applicability: Power 4)pe:'ation,.S.' Hot Stan.,b., Hotut.own ., 1,2. 3, 4 Action Levels: (1 ef-2-or 2 or 3)Note: The Emergency Director should declare the UNUSUAL EVENTUn-usual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(I)J RCS unidentified or pressure boundary leakage greater than (site speeifie "alu) 10_apý m for 15 minutes or longer.(2) RCS identified leakage greater than (site specific value) 2 for 15 minutes or longer.(3) Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer.Basis: This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.EAL #1 and EAL #2 are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). EAL #3 addresses a RCS mass loss caused by an UNISOLABLE leak tgrough an interfacing system. These EALs thus apply to leakage into the containment, a secondar'-side system (e.g., steam generator tube leakage-in-a P'R) or a location outside of containment. The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). EAL #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage..The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. Fe- PWRs. a AnaAo emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated). For ,,WR.. a skuck open Relief Valve (SRV) or leaka, .e is HO. *. .n. idel.ed either ide. iti.ed or til identi'fied leakage -y Technical Sp ecification s .nd thcrcforc.e, is not .applicable to I.., .....The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.171 1 .Pa. ae Escalation of the emergency lev-el via ICs of Recognition Category A-R or F.EMERGENCY CLASSIFICATION LEVEL would be SU4: EAL-1 Selection Basis: The STP Technical Specifications limit for tinidentified leakage from the RCS is 1 gpim. NEI 99-01 Rev.6 states to use the higher of the Technical Specification limit or 10 amm.SU4: EAL-2 Selection Basis: The STP Technical Specifications limit for identified leakage from the RCS is 10 gpm. NEI 99-01 Rev. 6 reQluirements are to use the hidher of the Technical Specification limit or 25 -im.SU4: EA.L-3 Selection Basis: The STP Technical Specification limit for primary-to-secondary leakage is 150 gallons per day through any one steam generator, but the specification does not specify the type of leakage. Therefore, STPEGS will use the leakage outside containment: which may include SG Tube Leakage, at 25 gpmi for 15 minutes or longer in accordance with NEI 99-01 Rev. 6 guidance.EFERENCES: I .STP Technical Specification Section 3.4.6.2 Reactor Coolant System Operational Leakage.Developer Notes:...AL 1 ..For.the Site specific. leak rate value., enter th. higher- o 10 gpm or the value specified in the site's Technical Spccifieations, for- this type Ef leakaige.EAL ft2 For the site specifie leaki rate value, enter the higher of 25 gpm or the value specified in the site's Technical Specific-ations for; this type of leakage-.For- sites that have Techinical Specifications flhat do not specify, a leakage type f~or steam generator tubedeveloper-s shol.d inclde an EiAL. for tube leakage gr.at.er thani 25 gpm for 1 minutes or-...C. .....Assi..ment Attributes:

3. i. o.172 1 P aa e SU5 ECL: Natifieati, on of Unuusal Ev ei-t"UNUSUAL EVENT Initiating Condition:

Automatic or manual (trip [PArR] 1' ,cram [BWR]) fails to shutdown the reactor.Operating Mode Applicability: PPAWef4 aei--1_ 2 Net: A manu.al a.tion is any eperatei action, or. set of actieiis, which cau th.e. ,itf i.d., to bef apidly Mserted int. the coe, and does n .t in.lud. Manually dri 'ing in control rod1 Or .im.plem.ntatin cf bo.ro inj..ti.n zrategies. Action Levels: (1 or 2)Note: A manual action is ANY operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategZies. (1) a. An automatic (trip [P :RI ' sram

• did not shutdown the reactor.AND b. A subsequent manual action taken at the reactor control eensek--Spanels is successful in shutting down the reactor.(2) a. A manual trip (t rp-E ,PR]AND b. EITHER of the following:

rB'WR],rdid not shutdown the reactor.IA subsequent manual action taken at the reactor control ee*+slesgpanels is successful in shutting down the reactor.OR 2. A subsequent automatic (trip [1WR] /' [BWR]) is successful in shutting down the reactor.Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip fp-P'4ý ..... [rI that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control eonselespanels or an automatic (trip [PWR] / srani [,. -R]) is successful in shutting 1731 Pa ye down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.Following the failure on an automatic reactor (trip-[P, WR] / scram [BWL R), operators will promptly initiate manual actions at the reactor control ce-iseleepanels to shutdown the reactor (e.g., initiate a manual reactor (trip [PWRI / scran [BW-r)). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.If an initial manual reactor (trip [P..R] / scram [9R]) -is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control eenseleePanels to shut down the reactor (e.g., initiate a manual reactor (trip / scram [BWRj)) using a different switch). Depending upon several factors, the initial or subsequent effort to manually ktrip+P-WR4--sc-fa-4BW-R})- the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor (trip [12WR] .,,,,...{-gWAR]signal. If a subsequent manual or automatic (trip [PWR , ý scram [BWR-) is successful in Shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.A manual action at the reactor control *eonsoelspanels-is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual (trip ["'W1]4 / serail-BWRj)). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control- eaise4espanels". Taking the MadIe Switch to SI, TDO\',Th is considered to be a manual s.ram action. [B1jR]The plant response to the failure of an automatic or manual reactor (trip [P, -R] / scram [.W.R]) will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control eoeolespanels are also unsuccessful in shutting down the reactor, then the emergency classification level EMERGENCY CLASSIFICATION LEVEL will escalate to an A4ei4-ALERT via IC SA5. Depending upon the plant response, escalation is also possible via IC FA I. Absent the plant conditions needed to meet either IC SA5 or FA 1, an UnustlaI'GetI-UNUSUAL EVENT declaration is appropriate for this event.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.Should a reactor (trip [PWRI / scram [.... R) signal be generated as a result of plant work (e.g.. RPS setpoint testing), the following classification guidance should be applied.If the signal causes a plant transient that should have included an automatic (trip-{PUP [a / si...._W-R]) and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated. If the signal does not cause a plant transient and the (trip [PWR] / Lcram [ R). faiure is detenrined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted. SU5: EAL-I, EAIL-2 Selection Basis: N/A 1741 P ag e

REFERENCES:

1. OPOP03-ZG-0004.

Rev. 45. Reactor Startup 2. 0POP03-ZG-0005. Rev. 86, Plant Stailup to 100%Developer Notes: This WC is applicable in any Mode in whichl the actual r-eactor-pcwcf le,'el could exedthe powerý leve a whi"ch the react.. is sh--utdown. A PWR with a Sh..td.....c....... level that is less, than. or ecgual te the rvetacto power- level whichi defines the le,'c boud oIPwer- Operation (Mode 1) will nieed to i:nelude Star'tup (Mode 2) in the Ope.atin.g Mode.Applic.ability. For if the ..eac.tor is, considered tO be ShutdWownV at Y% and Power- 1 59 Operattioii starts at >50%.. theni the IC is also applicable in Star-tup Developesr. maiy include site specific.FOP oriter"ia indicative of. a s.ee.ssf.l r.eactor shut.OW. ini ail-I statemen"t. the 6aSsll or both A..a reactor power leve)The trm; 'reaetor co~ntrol cnle"may b@ replaced Ewith the appropriate site specific term (eg. mai ealtel beaf~s4)EC-L sgnetArius:3L. 175 1 P a g- e SU6 ECL: Notification of Unusual Event UNUSUAL EVENT Initiating Condition: Loss of aIALL onsite or offsite communications capabilities. Operating Mode Applicability: P..i... Vc....., ht .......... , At i.Wtd.w. 1, 2... 3 4 1Fimple-Emergency Action Levels: (1 or 2 or 3)(1) Loss of ALL of the following onsite communication methods listed in Table S2.(site speciic list of communiication's mcthods)(2) Loss of ALL of the following Offsite Response Orga nization (ORO)ORO communications methods listed in Table S2.(site specific list of f"o.m..i..tio. methods)(3) Loss of ALL of the following NRC communications methods listed in Table S2.(-ste specific list ." c........ tion. .. ... .z) line Table S2: Communications Methods EAL-1 EAL-2 EA L-3 ONSITE ORO NRC* Plant PA svstem X o Plant Radios X" Plant telephone system X X X* Satellite phones X X* Direct line from Control Rooms to Bay X X Citv* Microwave Lines to Houston X X* Security radio to Matagordca County X* Dedicated Ring-down lines X" ENS line X Basis: This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.1761 Pa g e This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).EAL #1- addresses a total loss of the communications methods used in support of routine plant operations. EAL #2- addresses a total loss of the cominunications methods used to notify all OROs of an emergency declaration. The OROs referred to here are (see Developer Notes)Matagorda County Sheriffs Office. and Texas Department of Public Safety Disaster District in Pierce.-EAL #3- addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. SU6: EAL-1, EAL-2, EAL-3 Selection Basis: Lines not included for of'tfite communications to ORO and NRC included links that would need relaving of infornation. Links were obtained from procedures OPGP05-ZV-001 I. Elmereencv Communications.

REFERENCES:

I. OPGP05-ZV-00

1. 1, Emergency Communications Developer-Notes: 9.. ^ 1 The "site speci..c list of co..u.ni..

tio m.etlhods" .-.uIld ., 11 inOmudo ll LlniEationS mlethodS used for r ...inc plan ,mun, tin (e.g,., .....reial O .Site telep..n.S. Page party sysems. !radios'.etc.). This listing should include installed Plant e..uipment and components, and ... i...o.. and maintainied by. indi'.idupls. 161 EAL ff2 The "site specifi list of communications methods" should includ!Ee all eommunlications m.eth.ds u.sed to pe....rf .. initial emer.enc. n .... --...ati. to ORC. s as described in the site Emerzgency Plan. The liSting shOu~ld include installed planit equkipmen~t and eomponents. and! not items owned And maintained by indiv'iditals. Gsaamplic methods aire indondic tedtlephone lines, cmercain l telephone lines, radies, .atel lite telephon.es and inter.ct based communications technology. declaration. f+611m the COnRol" RoomE in aCordane w ...itht site Emer....y Pa and tlpiSalt .it.i ea EAl., f3 The "site speciflc list inicatiOns Shold inlude all .omL i atiO mlethod)ES used to perfOnaR inlitial emer-gency notificatiOns to the NRC as dese-ribed in the site Emergencey Plan. The listing Shuld.. include ins;talled plant equiipment and components. and not item ; s owned and mainaitainied by individuals. These methods ar-e iypieally the dedicated E~mer-gency NotIificationl Systemi (ENS) telephone liine and commnercial telephone lines-.ECL Assignnment Attributes:

3. 1.l.C 1771 Pagee SU7 ECL: Notification of Un'usual Event UNUSUAL EVENT Initiating Condition:

Failure to isolate containment or loss of containment pressure control.f{4Pw'R Operating Mode Applicability: PVwe pe-iatkS a:upetp, -a-nl-l -utdoewn 1. 2.3, 4 Exampie Emergency Action Levels: (1 or 2)(1) a. Failure of containment to isolate when required by an actuation signal.AND b. ALL required penetrations are not isolatede4Osed within 15 minutes of the actuation signal.(2) a. Containment pressure greater than (54e .pecie .5 , .sig.AND b. than .ne .ill t"ain of (site p. efi system or equipment) No Containment Sprayeentainment spra-y, rain is operating per design for 15 minutes or longer.Basis: This IC addresses a failure of one or more containment penetrations to automatically isolate (elese) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.F-e+-EAL #1-_; the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status -isolated or not isolated -should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.EAL #2- addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser

Fans) are either lost or performing in a degraded manner.178 1Pa e 4-1-3-This event would escalate to a Site Area Emergency SITE AREA EMERGENCY in accordance with IC FSI if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.SU7: EAL-1 Selection Basis: N/A SM7: EAL-2 Selection Basis: If containment pressure reaches 9.5 psig, Containment Spray will actuate. If no train of Containment Snrav is otneratine ner desicm, the ability to lower containmaent nressure is contnromised.

One train of Containment Spray (Technical Specifications 3/4.6.2) is defined as one containment spray system capable of taking a suction from the RWST and transferring suction to the containment sump.

REFERENCES:

I. OPOP05-E()-FOO5, Rev. 1. Containment Critical Safety Function Status Tree_,. OPOP05-EO-FRZI, Rcv. 9, Response to Hligh Containment Pressure 3. Technical Specifications 3/4.6.2 9. Devel~oper-Notes:.10. En1ter thle "Cite specific Pr-essure" alethlat actuates containimenit pressuire c0nt4ol SyStemAS (e.g., containmfent spray). Also enter the site specific containmenit pressm-e control s'ste'equip:--nt that sh-ould be operatfi-g per design if the conetain-men't pressure actuation sepon i eahed. if desir-ed, specifle eondition ind ications suchl as parameter, 9'lues call a be-efl ered (e.g., a containment spray flow rate le:;s than a ceeiain value).1. ELAL A2 is not app icable to th U.S. Evolu ary Power ReaCtIr (EPR) design.1.2.ECLAs--4gmueiit Attrihu~tc,. -3.1 .!.A.Is, 1791 Pa g e SA1 ECL: Aei4 ALERT Initiating Condition: Loss of a-4ALL but one AC power source to emergency buses for 15 minutes or longer.Operating Mode Applicability: o r Oper.tion, Sta-rtup. ..ot Standby, Hat S,,.... HtS d 3, 4 Fhitntpte-Emnergency Action Levels: Note: The Emergency Director should declare the A4ei:tALERT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. AC power capability to (site specifi. .y buses) ,aALL three 4160V4164 AC ESF Buses is reduced to a single power source for 15 minutes or longer.AND b. A-NVANY additional single power source failure will result in a loss of aRALL AC power to SAFETY SYSTEMS.Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU 1.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below* -.-A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator). 0---A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being baE-k-fed from the unit main generator. 2-A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being bac-i-fed from an onsite or offsite power source.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.Escalation of the emergency cassification level EMERGENCY CLASSIFICATION LEVEL would be via IC SSI.1801 iPa P ge SAI: EAL-I Selection Basis: This EAL is similar to IC C(.2, except this EAL applies only to Modes 1-4.

REFERENCES:

1. OPOP04-AE-000 1 Rev. 44, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus.OPOP04-AE-0004.

Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus 3. OPSP03-EA-0002. Rev. 32. ESF Power Availability

4. DrawinQ OOOOOEOAAAA.

Rev. 24, Sinale Line Diagram. Main One Line Diagram. Unit No. I& 2 Developer Notes: For a power source that has nEltiple the ý and.,or Basis section should reflect the minhium number of operating geiratcrs necessary I;"r that s....ec to provide r-equired pow. .. to an AG.... ........... .. .... ..... .. ... q....... ." .. .lle ý,lc ý -po e; -o : o ......ýenerators fori that source arc oper-ating. The "site specific emlergency bu!Ses" are the buses fed bN' offe-ite or emerfgencyf ~AG power- sources thlat Supply ..O.e.. to the ejLec.trical dis.ribution y,, ,stem' that powerfes SAFETY SYSTEINS. There i5 tvpiall.;enmergency bus per train of SAFETY SYSTE.MS.De'se!OperS should modify the bulleted examiples proevided iin the basis section, above, as needed to refl their site specific Plant designs and capa:ilities. The .,ALs and Basis s.ul reflect tha ech indepen.dent. f..ite po.A constitutes a single PO.sour.e. FIr e.ample, thee ind ..ependent '3451NV a.i.e powe* -.ircuits (i.e., incing .power lin.es)cri-ps 'fee se power souces.. .nependenee ay-be leteRm4fne-i-"u! a reVie-v of the site-specific UFSAR, SBO analysis or related loss of electrical powe'r studies.Tile EAL and,'or Basis section may. speci',' use of a noni safety r-elae power- SOurce provided tha opelatian of this source is r.....i..d in A01l Ps. or beyond desion basis acdn re&, e s.guidelines (e.g., FLEX suppot guidelines). Suc-h power SOurc'es shou..ld gen..rally mieet the source" definition provided in 10 CER 50.2.At m.ulti uit StatiOnS, the EALS MW,'ay credit comlpensator,, me..asres tat are proceduralized and cani be ipm ted within I5 mIntes. Consider capabilities such, as power-source c.o...ies. , "Swing" generFatorohe! oe sources dlescribed in abnorm-nal or emergency. opr Fee poedures, etc. Planits that have a proceduralized capability to suppl-y offsite AC power o -an-a " .... sis ECo a may c-redit this, power soue in tile AL provided that th.e plannied cross tip.tr mieets the reureet o 10 CER 50.63 EG-L Aessizment Attributes:i 3.1 .2 4 1811 Pa R e SA2 ECL: Alei4 ALERT Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.Operating Mode Applicability: Pwer Opeation, Suita-ip, Hoet Sta-dby. Hot Shutdown-, 2, 3.4 E-aniple-Emergency Action Levels: Note: The Emergency Director should declare the Ale4-ALERT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. An UNPLANNED event results in the inability to monitor one or more of the following parameters in Table SI from within the Control Room for 15 minutes or longer.[B JJ'R AW]zcs 's [P U';' pr','e~ef lkit]Reaceter Power Reaeter Pcwcr RPV Water Lev~el RGS Lev~el RPV Pressure RGS Presseue Primar-y Containment Pressue in Ex:it Temperature Suppression Peal Leve Levels in at least (site specific niumber-) two steamf generators Suppr-ession Pool Temperaturie Steam Generator Auxiliafy--or Emergency Feed Water: Fl&w Table SI: Plant Parameters

• Reactor Power" RCS Level* RCS Pressure* Core Exit Temperature
• Levels in at least two steam generators
• Steam Generator Auxiliary Feed Water Flow AND b. ANY of the following transient events in progress." Automatic or manual runback greater than 25% thermal reactor power" Electrical load rejection greater than 25% full electrical load 1821Page 0 Reactor seram rBIA ./-W /trip+P-W-j 0 ECCS (SI) actuation Thr-ma .....v[o-'r- oscillaqtionIs e-a~ter titaIno/ 1",4, lB Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room.During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s).

For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling LrD fI] ,' RPV level [B"r] and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reat.r !RCS level [P,,J2] / RPV water., ... , rIPW cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the e'nergency classification level EMERGENCY CLASSIFICATION LEVEL would be via ICs FS I or IC A-S-.RS 1.SA2: EAL-I Selection Criteria: The plant parameters listed are fiom NEl 99-01. Rev. 6. Two steam Penerators were selected as a site-specific parameter for the minimum number of steam generators needed for plant cooldown and shutdown.

REFERENCES:

I. OPOP05-EO-EO20. Rev. 11, Faulted Steam Generator Isolation 2. OPOP05-EO-FRHI. Rev. 23. Response to Loss of Secondary Heat Sink 183 P a , e Developer-Notes.!ill tile PWR paramleter list eaol~iln. the "Site specific number', shoukld reflect thle miiuuber of steaml genieiataps niecessar-y for plant eooldowN~ anid shuitdown. This critierion mla", alIso specify, whether. tile level value should be- -'Ae-r ,n "a'ge or both, d ipm-r t Dev'elope!rs may speeify either: pressurizeri A-;. reaetoir vessell levell i.4 the PWR paramneter! EOlvMn4 entr-y for.Developars should consider, if the "tr'ansienit evenits" list needs to be Modified to better' reflect Site specific plant operating, char-acteristic-s WWI expected respon~ses. The nu~mber, type, location an d layout of Control ROOM ind~icattions, and the rangz-e Of possible faýilurfe miodes, can challenge the ability of an operator-to accurately deter'mine, within the timie per-iod av.ailable fo;r P~emergecy classifleatiosi assessmenits., if a specific perceentage of indicationis have been lest. The appr.oach used ini ,his EAL .acilitates prompt and accurate meirgene' ass<:essments by faeusin'" -I-, the inidications f-or a selected subset of parameter's. By f.cusi-. g on t.ie availability of the specified paramieter Values, instead4of the sources of those k1a.les..the EAL recognizes and accommodates the wide va.i. e of i.ndiat.. ions in nuclear. power. plant Control REooms.n Indication ty-pes and sourcees miay be anialog or digital, safety related Of not, pr~imnal,N or alternate.,ý-tn&4tual .-teF ake CIptter &FO;.ip i~slt-aý-',-ete. 168 A loss of plant annuncfe~iators "ill be ev'aluated for reporlab.i1iyi ccrac it40CR 07-a.......... ....................... ,. ...... .. ..t... .... w : c o~a c !0 CFPR 50.72 (a.9 the associated guidance in NtUREG 1022). and repor'ed if it sinifi cantly impairs the capability to perform+... emerg.ency assessments. Compensat...' ..easures far a loss of annunleiation can be readily implemented and may, include increased monitoing of main conrol boars an.d maore frequent plant roun..ds, by non licensed operators. Their alerting ...function no.twithstand.e. ......an.nun.iators do fat pr'vide the par.a.eter values or- speeific component stats infrm-ation used to operiate the plait, o. process through AOPs or EOPs. Based Em these conisider-ations, a loss of alnnunciationi is consider-ed tob adequbately addressed by' rePOrtability' criter-ia, anld therefqire not included in this IC and EAL.With respect to establishing event severity, the response to a less ofFd--l process or effluent monitor values) is to be adequately bounded by the requirements of 10 CER 50.72 (and associated ,o 11.. 1e1 ...n.. 1- "NREi 10212.1' The r.epor.ting of this e.enlt will ensure adequate plant staff an.d .RC awareness, iand drive the establis.ent ofappro prate compensat0o:' measur.es and corrective actions. in addition, a loss of r-adiation monitor;ing data, by itself, is not a pr'ecur-sor to a mor'e s:infi~eanit evenit.Per)solnnel at siteS that haVe a Failur,-e Modes anid Effects Analysis (PAMEA) included Nkithin the , basis- ofaRdigital !&-C sy'stemi Should consider-the EN'IEA ki forme44efiwhenl developing their Site speceific. ue, to changes in thle con.fig uation.s of SAFETV SYSTEMS, i..,..din.g associated instrumentation ad idiations. during the cold shutdown, refueling, and defucled modes, no C- is incuded for these modes' orfoper-ation.7 4 1841 Page SA5 ECL: A#ei4-ALERT Initiating Condition: Automatic or manual (trip [PWR] /ý s;eam [.BWR]) fails to Shutdown the reactor, and subsequent manual actions taken at the reactor controleen-eose panelscontrel eonso are not successful in shutting down the reactor.Operating Mode Applicability: Pewe Opei-at4ionl. 2 Note: A manual actioni is any operator action, of set of aetiens, whieh causes the control rads to bcre rapidly inseited into she core, and does not include mfanually dr-ivinig in control rceds or implemenitation oa boron injection strategies. Emergency Action Levels: Note: A manual action is ANY operator action. or set of actions. which causes the control rods to be rapidly inserted into the core. and does not include manually driving in control rods or implementation of boron injection strategies. (1) a. An automatic or manual (trip [PWR] / scram [BWR) did not shutdown thle reactor.AND b. Manual actions taken at the reactor control eonselespancls are not successful in shutting down the reactor.Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip [-PW-R/-sera-{-8BWRI} that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control ee-sel-,e&panels to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control eensoe-panels since this event entails a significant failure of the RPS.A manual action at the reactor control c-onsetespanels is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor (trip-P-W. rI). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control el (e.g., locally opening breakers). Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control eefinlespanels". Taking, the Reactor. Mo.de Switch, to SHUTDOWN is ..nsider;ed to be a m ,anual scramfl atiefn. [1WR.J The plant response to the failure of an automatic or manual reactor (trip [PWR] 1 scram [14N[trR]) will vary based upon several factors including the reactor power level prior to tile event, availability of the 185 Page condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If tile failure to shutdown the reactor is prolonged enough to cause a challenge to the core cooling j4WR+ /R4W w.,ater, le.e'il [BWR.] or RCS heat removal safety functions, the emergency ,lassifiation level EMERGENCY CLASSIFICATION LEVEL will escalate to a Site Area Emnergency SITE AREA EMERGENCY via IC SS5. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant ,.nditi.ns needed to meet either , C or PSI, an Alert de.larati.. i;agprpri-atc foar this event.It is recognized that plant responses or symptoms may also require an Alei- ALERT declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration. A reactor shutdown is detennined in accordance with applicable Emergency Operating Procedure criteria.SA5: EAL-I Selection Basis: N/A

REFERENCES:

1. OPOP05-EO-FRSI.

Rev. 17, Response to Nuclear Power Generation -ATWS Developer-Notes: Thi IC is applicable in any Mode in. whi.h the actual r.eactr. p-wer level ..uld exceed the powe. level a4 whici the iS co.Sie... d .ihutd.wn. A IPWR ,ith a r..etor. pOwer; lev.el that is less thani .r equal to the@ raeator power level whichi defines the lower bound of Powher Oper-ation (Mode I) will need toStartup (Mode 2) in the Operating Mode Applicability. For- example, if the reactor is considered to he Shutdown at 39% anld Power Operation StartS ait >5,%. then the K; iS alsO applicable. ini Startup Mode.Developers mnay include site speei4ic-EOP cr-iter-ia indieative of a successiful r-eactor Shubtdow.n in an EAL1.statement, the Basis or both (e.g., a reactor power level): The term "r.eactor .otrl consoles" may. be replaced with the appopriate site specific term... (e.g.,-eointfol b ards).EC-L Asisignmfenl A i,.tlv; ,. i 2 1 '1 i I 1861 P ag e SA9 ECL: Alet4-ALERT Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.Operating Mode Applicability: Power Operation, Sta.tup.. Hot Standby. Hot Shutdown-1, , 3. 4-x-ampte-Emergency Action Levels: (1) a. The occurrence of ANY of the following hazardous events listed in Table S3: Saeismic even (~earthqak i ~e) en High winds or tornado strike F4RE (site specil hazards) Pr-edicted or actual breach of Main Cooling9 Reservoir retaining dike along North Other ev.ents with similar hazard char-acteristics as determinied by the Shift Managef Table S3: Hazardous Events" Seismic event (earthquake)" Internal or external floodiný event" Hi-h winds or tornado strike" FIRE" EXPLOSION o Predicted or actual breach of Main Cooling Reservoir retainin2 dike along North Wall.o Other events with similar hazard characteristics as determined by the Shift Manager AND b. EITHER of the following:

1. Event damage has caused indications of degraded performnance in at least one train of a SAFETY SYSTEM needed for the current operating mode.OR 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.187 1 P a g e Basis: This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.EAL# L.b. l- addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.

The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.EAL# l.b.2- addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.Escalation of the emefgef'ey via IC FSI or A-RSIAS4-. elassifieation levcl EMERGENCY CLASSIFICATION LEVEL would be SA9: EAL-1 Selection Basis: The listed hazards are from NEI 99-01 Rev.6 with the exception of the Main Cooling Reservoir breach alon~g the north wall which was included because it is a credible hazard and analyzed in the STPEGS UFSAR.

REFERENCES:

I. STPEGS U FSAR, Section 3.4.1. Flood Protection Dcvelopcr Notes: For- (site spceifi hazafds), Edevelepecrs should eansidef ineluding, e~hei sigiiific-ani, site speei fie haixaids te the- bulleted list c-ontained in EAL l.a (e.g., a seiehe).>iuclear powAer Planit SAFE.Y SY&STEM-NS are C~OMPri~ed Oftv.'o at- mor-e separate and redusidant traiins El ECL zinetAri 1~ 1 1)D.P 188 Pa R e SS1 ECL: Site Area Emergency SITE AREA EMERGENCY Initiating Condition: Loss of IIALL offsite and aRALL onsite AC power to emergency buses for 15 m11 inutes or longer.I Operating Mode Applicability: P&" ei- Of eration-sta!4up' l= 't S+--- byn 14- Ot ShE!!--1.2. 34 I*-v-plEmergency Action Levels: Note: The Emergency Director should declare the Site Area Emergency SITESITE AREA EMERGENCY promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Loss of ALL offsite AND ALL onsite AC power to (site specific emergency b ) ALL three 4160V AC ESF Buses for 15 minutes or longer.Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emeFge+-ey el-via ICs 1, FG 1 or SG I.SS I ;I r AP All leVel-EMERGENCY CLASSIFICATION LEVEL would be SSI: EAL-1 Selection Criteria: N/A SS4. EA, 1

REFERENCES:

I. OPOP04-AE-0001. Rev. 44. First Response to Loss of Any or All 13.8 KV or 4.16 KV BUs 2. OPOP04-AE-0004. Rev. 15. Loss of Power to One or More 4.16 KV ESF Bus 3 OPSP3-EA-0002, ReN. 32. .ESF Power Availability

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Rev. 24, Single Line Diauram, Main One Line Diagram, Unit No. I 1891 Pa a e Dele'cOpeF Notes!For a t)8NN'er Source that has mulltiple gnearsth EAL and/oer Basis sectioii should refleet the, minimuminuber f pE4ain gener.1-1 -ators !Iecessary for- that SOurcee to prFOvide adequate power to anA gtecy bus. For example, if a bae !p power ... r.. is comprised of two generators (i.e.. t 0o 50%eapaEity .gei....O.S SIed t. fed , AC_ emergenyu, th... [EA!.., and ,asi; seetion must speEify that b'gener'ato"s for that source are operating. The "'site specifie emergeney buses"' are the buises f-ed by ite or- emnergeney AC power soutrces that supply pwErN-@ to the electrical distr-ibutiOnl Sysltemf that poIAer's SAFETY SYSTEMS. There is typically emnergency bus per tr-ain of SAFETY SYSTEMS.The EAL and/,or Basis section may specify, use of a nion safety related power source provided ta prain of this source ii cnitr-olled ini accordancee with abormeiial or em~ergency operatinig proceduires, beyon design basis accident response guidelinies (e.g., FLEX support guidelines). Suchi power sources shud eerally meet the "Alterniate ac source" definition provided in 10 CFR 5-0.2.Aýt muclti unit stationis, the IE -A 6s may; ercdit eomnpensator:, measur!esl that ar. procedufali Zed anid ean be ifimpl emented within 15-) minlutes. Con4siderF capabilities such as power source@ cross ties, "swinig" ecaratrsoter--wi' sourcees deseribed in abnormHal or emergenciy ortigrcereset. Plat tat h ave a preC-e dural ize d ca p abilty to suLlp ply) A ffi le ACG poWer- to E)I-I affecte d unt vi a E.os -1 to, a copaio unit na:, credit thiS pow"er source inl thle FAL, provided tat the plannged croass tie strategy m~eet: the requirements of 10 CER 50.6-3.OF 1-IZLAssignmlenit A0419i 1 itt-,; ý -I 1 -4 m 1901 P a g e SS5 ECL: Site Area Emergency SITE SITE AREA EMERGENCY Initiating Condition: Inability to shutdown the reactor causing a challenge to (core coolingrP,-AR+, RP,,V water levl ") or RCS heat removal.Operating Mode Applicability: Pewei-pie+i- _I2 EFxampt e Emergency Action Levels: (1) a. An automatic or manual i-trip [PWR] / szcram[BAV]) did not shutdown the reactor.AND b. AIIALL manual actions to shutdown the reactor have been unsuccessful. AND c. EITHER of the following conditions exists:..i.. .7t,-4e1 i i c at.. ...f ..n inabilit to ade...atel. , Cooling -Red entry conditionseenditien met OR i~emov3\e heat fr--m the car-e) Core_... ; --pecifie i,.diatin of.an i, abi Red entry conditionseen.d.i.ti... met lity to adequately rello','e heat fi'em the RCS) Heat Sink-Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip {-P-R-' Leram [B4]R+that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency SITESITE AREA EMERGENCY. In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency SITESITE AREA EMERGENCY in response to prolonged failure to shutdown the reactor.A reactor shutdown is detenrined in accordance with applicable Emergency Operating Procedure criteria.191 Page Escalation of the emergency le-el EMERGENCY CLASSIFICATION LEVEL would be via IC ARG I AG--RG I or FG 1.SS5: .EAL-I Selection Basis: Core Cooling -Red entry conditions met (CETs > 1200' F) is the site specific indication of the inability to adequately remove heat from the core. Heat Sink -Red entry conditions met (NR level in All SG <14% [34%] AND total AFW flow to SG < 576 GPM) is the site specific indication of the inability to remove heat from the RCS.SS4: EA4,4-REFEIRENCES:

1. Procedure 0POP05-EO-F002, Rev. 2. Core Cooling Critical Safety Function Status Tree 2. Procedure OPOPO5-EO-F003, Rev. 6. Heat Sink Critical Safety Function Status Tree Deve'cepeo" Notes: This 1C is appi4eab4e in an'; Mode in which the actu-al reactor power level could exeeed the power level at whichi tile reactor is conisidered shutdown.

A PWRwt htonratrpwrlvlta sls hno equal to dhe rwector poA'.er level whichi defnes the lower bound of'Power Opetation (Mode 1) will need to in.lu.de Sta.tup (Mode 2) in tie Operati. g Mode .A.pplicabi!it'. For ex.ample, if the reactor. is considered Devclopers miay include site specific BOP cr-iteria inldicativ~e of a suaeeessful reaetor- shutdown in an EAL staten ent. the Basis or both (e.g.. a Ie-a...- powerl.eve4 Site specific inidication. of an inability to adequately .em.ve heat fro. m the .....R.. .Reactor .vessel water lev..el cannot be res.ored and maintained above Mini.mm Steam Colin.RPNV Water Level (as descrFibed in thie EOP bases').[PFlU] hIsert site specic ,i"fie l fo. r9 an; .n.icor./co.e. exit ther-nmocouple tem.perature and/oer. reaetor vessel waer, level that dives entry into a cor.e cooling restor.ation proedre (or otherwise ru...e.s implementation of prom..pt r.estorationi actions). Al .at may ise ineore/co'e exit t.erm...upl.e temp.ei'a-'es' atd-1.',r a reactor vessel water level th e

• e sepeds4f--afepa.

e-the middle of active ruel. Plants with vessel level i..s.umentati h.at cannot imeasre down t appoxmatlythe middle ofaetive fuwl should use the lowest on scale nidn hat sis n bo.e the topof aetv..e f.el. lithe lowest. scale readin. is abov.. e the top of a.tive fuel, then a ..ea. .v e level val-e should not he inwacluded. Feorlat that have implemented Westinhous ....e Ownes Group E e.. Respon.se Guidelin.es. enter the parametes .ed inl. the Core Cooling Red Path.Site specific indication. fi a inability to adequ.a.ely ove heat froi the RCS:[B9PPW] Use the Heat GCf_-,aeit. Temper'ature Limit. This addresses the inability to remove heat via the maain condenser-and the suppression pool due to high pool water- temperature."[PIIR] Insert site ,pecific Paramctei's assoc-iated w0ith iniadequate RCS heat i'emoval via the steam generators. These parameters shoulId be identical to those used for' the I!adequate Heat Removal thireshold Futel Clad Barreie Potential Loss 2.13 anid thr-eshold RGCS BarrFier Potential Loss 2.A in the PWIR EAL Fission ProducIt Barrier Table.ECL- Assignment Attributes: 3.1 ..B-192 P a g e SS8 ECL: Site Area Emergency -SITE AREA EMERGENCY Initiating Condition: Loss of a-4ALL Vital DC power for 15 minutes or longer.Operating Mode Applicability: Power Operation, Hot Standby, Heot Shut down L., 3 4 Ea-mple--Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergency SITE AREA EMERGENCY promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Indicated voltage is less than (site specific bus v.ltage val.) "105.- on ALL ( s4+e-speifie DC b".sses) Class IE 125; VDC battery buses for 15 minutes or longer.Basis: This IC addresses a loss of Vital DC power which compromises tile ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level EMERGENCY CLASSIFICATION LEVEL would be via ICs AG-IRG 1, FG I or SG8.SS8: EAL-1 Selection Basis: Minimum voltage for Class I E 125 VDC batterv buses was determined in calculation 13-DJ-006 Rev.3 and determined to be 105.5 volts. At 105.5 volts or less, 0POP05-E0-EC0O, Loss of All AC Power directs the operators to open the battery output breakers.

REFERENCES:

1. OPOP05-E0-EC00, Rev. 23. Loss of All AC Power DpevelprNotes.!

The"site specific bus vol.age value" should be based on the .minimum bus v"oltage necessary for adequate oper-ation of 9,AFETY) SYSTF=N4 eq~iipmfent. This v.oltage value sl ld iHCEeFPor-ate a nmaran ofat least 15 oinutesooperatioen before thie onset of.. 'bility to operate these leads. This, vltage isusual ylieai-he mimmvakai~c selected wAhenl battery swill.- pelrfeflmed. Th ,,. valu. for an... entir batter; ....t i -app.......imately 105 VDC. Pei a 60 .ell.strin of batteries, the ,ell voltage i VU.....i.ately 1.75 Volts per c..l. F, r a 5 t attery set, the 1 i41iiiinimum voltage ,i1 app oximately 1.. 1 Volts per cell.The "site specific Vital DC busses" are the. DG busses that provide m.nitoring asid capabilities for.sSAFETY SYSTEMS.1931 P a g e I EC-6 AssIgnment Aitrihltptc' 21 131 1941 P a g e SG1 ECL: General genc......y GENE.GENERAL EMERGENCY Initiating Condition: Prolonged loss of a-4ALL offsite and a44ALL onsite AC power to emergency buses.Operating Mode Applicability: Power Operation-, Startup. Hot Standby. V!t Shutdown 1. 2. 3, 4 Sa*m+pek-Emergency Action Levels: Note: The Emergency Director should declare the General Emergency GENERALGENERAL EMERGENCY promptly upon determining that (hite specific how'6s) hours has been exceeded, or will likely be exceeded.(1) a. Loss of ALL offsite and ALL onsite AC power to (t"e spec-ific eme...enc.. -c'ALL three 4160V AC ESF Buses.AND b. EITHER of the following: " Restoration of at least one AC emergency 4160VAC ESF bus in less than (site specifie haEis-)-4 hours is not likely." (Site sPe ifie indiation of all iiability to adeqUat, , remo.ve heat from the core) Core Cooling- Red entry condition met Basis: This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. The EAL should require declaration of a Ei,,,g..... GENERALGENERAL EMERGENCY prior to meeting the thresholds for IC FG I. This will allow additional time for implementation of offsite protective actions.Escalation of the emergency classification from Site Area Emergency SITESITE AREA EMERGENCY will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of fo)ur (4) hou~irs-the awflyized stati.n blackotc eid... Beyond this time., plant responses and event 1951 Page trajectory are subject to greater uncertainty, and there is an ihefease4higher likelihood of challenges to multiple fission product barriers.The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.The EAL will also require a General EmergencyGENERAL EMERGENCY declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.SGI: EAL-1 Selection Basis: The prolonged loss of all onsite and all off site AC power coupled with Core Cooling -Red entry conditions (CEIs > 1200' F) are sufficient indications of the inability to remove heat from the core.Station Blackout does not include the loss of available AC power to buses fed by station batteries through inverters. or by Alternate AC (AAC) sources as defined in NUMARC 87-00. The STPEGS Station Blackout position credits any one. of the three Standby Diesel Generators as the AAC source. The required coping duration category determined for STPEGS Station Blackout is a minimum of four hours, based on the guidance of NUMARC 87-00. Section 3. STPEGS meets this requirement and forms the basis for the four hour time neriod.

REFERENCES:

I. OPOP04-AE-0001. Rev. 44, First Response to Loss of Any or All 13.8 KV or 4.16 .KV Bus 2. OPOP04-AE-0004. Rev. 15, Loss of Power to One of More 4.16 KV ESF Buses 3. OPSPO3-EA-0002. Rev. 32,.ESF Power Availability

4. Drawingy OOOOOEOAAAA.

Rev. 24. Single Line Diagram. Main One Line Diagram. Unit No. I 5. OPO.P05-EO-F002. Rev. 2, Core Cooling Critical Safetx' Function Status Tree 6. OPOP05-EO-ECOO, Rev. 23. Loss ofAll AC Power 7. STPEGS UFSAR. Section 8.3.4. Station Blackout Developef Notes:_.Althoulgh this WC and EAL may be kiewed as redundant to the Fission Product Barrier- !Gs. it is included to provide for a m"ore time; v es.alation of the .y. .lassification level.The "site specific emergen.y bu..ses. " are the buses fed b` off-ite o-. emner-gen--cy AC soulr-ces that suppl.'.. power the e@lectical distribution Systen, that pwo'-ers SAFETY SY-STE.MS. There is typically 1 emergency bus Pet: train o.f SAFETY SYSTEMS.-The "site specifc hour-s"; to estare AC poerit to an emfer~gency' bus shiould be based en thie statir blackout coping ..nal. ys performed in ac...rdance with 10 ..R, § 50.63 and Reaulator-y Guide 1. 155.stwNiei Bloo6out Sit-e-&peeifie indication of an,-4nabili, to adequately remove heat Irm til-oe re:[BJJ'R] Reaetei-,ese w.ater level cannot be restor-ed and maintaindA _abeole M44inimum Steam Cooling RPkV Water- Lcvel (as descr-ibed in thle 90P bases).[pJ-] insert' specific A/ exit th1armocouple te-+lpere tur " and/or reactor "esset Water le-el .hat `ri ..ntr into a cor.e .ooling restoatio.in praocedure (or othe-Wise r.equires 196 1 P a , e teiiperaturec. grcatei, than 6 200oF and.'E) a reaetor vessel wiater le-vel thaat corrcv~pondc t0 apliromiffately thle middle ofaetive fuel. Planits withi reactor- vessel level instirurentation 4hat cannot mieasure downl to a t is not aboe active fuel. if the lowkest ai seale r-eading abv th to faciefel, thn a arieactor vessel level value Tho-ud not be iRCeLIEWed. ForL plantsetat have iniplemcnted Westinghouse Owniiers Gr-oup Emer-geiEYR n Goidelies., enter the param -t-re -ed in thie C~ore Caookige Red Path.ECL Assi ..... -,,A ibt 3.1.4.9 197 1 P a g e SG8 ECL: General Emnergency GENERAL EMERGENCY Initiating Condition: Loss of a4ALL AC and Vital DC power sources for 15 minutes or longer.Operating Mode Applicability: Powes- Opei-ation, Sta.tup, Hat Staidb;.. ot, SHt. .. .34 Emergency Action Levels: Note: The Emergency Director should declare the General Emer'gency GENERAL EMERGENCY promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. Loss of ALL offsite and ALL onsite AC power to (site specifi- cmre:'genev buses) a!ALL three 4160V44460-- AC ESF buses for 15 minutes or longer.AND b. Indicated voltage is less than (site ......ific bus voltage ,,,tu) -105.5 VDCQols-P on ALLALL (site .pecifie Vital 1,C bws.) Class I F 125 VDC battery buses for 15 minutes or longer.Basis: This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.SG8: EAL-1 Selection Basis: This IC and EAL were included to address the operating experience for the March, 2011 accident at Fukushima Daiichi. Minimum voltao.)e for Class 1E 125 VDC battery buses was determined in calculation 13-DJ-006 R ev.3 and deternined to be 105.5 volts. At 105.5 volts or less. 0POP05-E0-ECO0, L..oss of All AC Power directs the operators to open the battery output breakers.

REFERENCES:

1. OPOP04-AF-0001, Rev. 44. First Response to L..,oss of Any or All 13.8 KV or 4.16 KV Bus 2. OPOP04-AE-0004.

Rev. 15. Loss of Power to One of More 4.16 KV LSF Buses 3. OPSP03-EA-0002. Rev. 32, ESF Power Availability

4. 0POP05-E0-ECOO, Rev. 23, Loss of All AC Power 198 1 P a g e

5. Drawing OOOOOEOAAAA, Rev. 24. Single Line Diaaram, Main One Line Diagram. Unit No. I&2 Developer" Note;: The "site specific emergency buses", ar-e the buses fed by; fiEmor emergency AG powetr sources, that supply power' to the electr.ical distribuktiai Sy.stem that power. s SAFETY SYSTEMS. There is tYpic a 1; I egei- y-bus-,-Per train of SAFETY S-YSP EI The "site specific voltage ..alue. sh.. ld be based oni the mini.lmum, b.. svoltage necessar.

y I:. .adequa.e operation o ETY SYS TE M eqSipment. This wltage value should incorporate a margin o. at least 15 pp~inate-s of operation before the onset of i nabil ity, to. operate those la&.hiS votg sUsually near- the minu4 t slected whet batter'y sizing is Peirfmed.The typieal value for an entire battery set is approximately 105 tDG. For a 60 eell strini of b tteri.s, the Oeli vo~ltage is ppr-iately 1.7-5 Volts per cell. Fur- a 53srn battery setemnmi ot+-*approxiatel, I .91 Volts-e-cl.. Tie "site specific %Vita.1 Dc- bulses" arce thle DC busses that provide moinitor-ing and controli capabilities fbir SAFETY SYSTEMS.This Wc anid EAI.. wer-e added to Revision 6 to aEddi".iatm _Xpeienee from thie M,'afh, 2011 aeccident at Fulaushima Daiichi.ECL6 Assiagnmient Attributes: 3.1.4.l3 199 1 P a ge APPENDIX A -ACRONYMS AND ABBREVIATIONS AC ................................................................................................................................. Alternating Current AOP ........................................................................................................... Abnormal Operating Procedure APPAII ............................................................................................................. o,-,erage Power- Range NM eter ATW S .............................................................................................. Anticipated Transient W ithout Scram B&W ........................................................................................................................... Babcck and W ilco-lUT ................................................................................................ Beron Injection Initiation Temper-ature BWBR ........................................................................................................................ Boiling B i W ater Reactor CDE ................................................................................................................ Committed Dose Equivalent CFR ................................................................................................................ Code of Federal Regulations CTM T/CNM T ......................................................................................................................... Containment CSF ........................................................................................................................ Critical Safety Function CSFST ................................................................................................. Critical Safety Function Status Tree DBA ......................................................................................................................... Design Basis Accident D C ......................................................................................................................................... D irect C u rrent EAL ...................................................................................................................... Emergency Action Level ECCS ...................................................................................................... Emergency Core Cooling System ECL ........................................................................................................... Emergency Classification Level EOF ............................................................................................................ Emergency Operations Facility EOP .......................................................................................................... Emergency Operating Procedure EPA ........................................................................................................ Environmental Protection Agency EPG .......................................................................................................... Emergency Procedure Guideline EPRG ........................................................................................... Ef....gell. .Pla ime plemienip nge Prucedure 91214 ......................................................................................................... Emergency M aagýP w emen t A en y EPRI .................................................................................................... Electc PFi lwer Research IsistitUte ERG ............................................................................................ Emergency Response Guideline FEM A ......................................................................................... Federal Emergency M anagement Agency FSAR .............................................................................................................. Final Safety Analysis Report GE ................................................................................. G .. ..l Em .eF, .. ...yGENERAL EM ERGENCY HCFSL ........................................................................................ ndependentpHeat CapacityF Temper atu aae imit Keff m ....................................................................................................... tive N -eut iron .M u -Altp lica tion aco 4S! ......................................................................................................... .... HumaC n Systemo n Ifnteraface IC A ................................................................................................................... .... Initiating Condition ID ................................................................................................................. ........ Inside Diameter IPEEE.............................. individuial Plant Examination of &Eternfal Ev~ents (Genieric Letter- 8.8 20)ISFSI ............................................................... Independent Spent Fuel Storage Installation Keff ..................................................................... Effective Neutron M ultiplication Factor LCO ...................................................................................................... Limiting Condition of Operation LOCA ................ ....................................................................... L ossR of Coolant Accident M CR .............................................................................................................................. m ain C-01trol1 ROOM M SIV ................................................................................. M ain Steamn Isolation Valve M SL ........................................................................................................................ Mar Stear m Line mR, mRemn. mrem, rnREM.....................................................mrilli-Roentgen Equivalent Man M W ......................................................................................................... M egawatt NEI ........................................................................................................................ Nuclear Energy Institute NPP..............................2Nuclear Power Plant------- --------- 200 1 P a R e N RC ......................................................................................................... N uclear Regulatory Com m ission N SSS ........................................................................................................... N uclear Steam Supply System NORA D ............................................................................ N orth Am erican Aerospace Defense Com m and (N O)UE ................................................................................................... (Notification Of) Unusual Event N UM ARCNU. .A RC.9......................................................... Nuclear M anagem ent and Resources Council OBE ................................................................................................................ Operating Basis Earthquake OCA ....................................................................................................................... Owner Controlled Area ODCM,/ D AN4 ............................................................... Offsite Dose Calculation_ ..ssessm ef.. .-M anual ORO ........................................................................................................... O ff-site Response Organization PA ........................................................................................................................................ Protected Area PACS ............................................................................................. Priorit)y Actuation and Control System PAG ................................................................................................................ Protective Action Guideline PICS ............................................................................................ Process inform ation and Contre System PRA/PSA ............................................... Probabilistic Risk Assessm ent / Probabilistic Safety Assessment PW R ................................................................................................................... Pressurized W ater Reactor PSIG ............................................................................................................ Pounds per Square Inch Gauge R .................................................................................................................................................... Roentgen RCC ..................................................................................................................... Reactor Control Consoe RCIC .......................................................................................................... Reactor Core. l5sol!tion Ced ing RCS ....................................................................................................................... Reactor Coolant System Rein, remn, REM ................................................................................................. Roentgen Equivalent M an RETS ................................................................................. Radiolo gical Effluent Technlcal Specifications RPS .................................................................................................................... Reactor Protection System RPV ........................................................................................................................ Reactor Pressure Vessel RVW L4.- .................................................................. Reactor Vessel W ater Level Instr.. en. at.. n System RW CU .................................................................................................................... Reaector W ater Cleanup SAR ......................................................................................................................... Safety Analysis Report SA9 .......................................................................................................................... ......... Station Blac.l. ut SCBA ................................................................................................. Self-Contained Breathing Apparatus SG ..................................................................................................................................... Steam Generator S1 ......................................................................................................................................... Safety inlection SICS Safety inform.ation and System SPDS ....................................................................................................... Safety Param eter D isplay System SRO ............................................................................................................. ....... .Seni or Rea tor Oper t TEDE ................................................................................................... Total Effective Dose Equivalent TOAF .......................................................................................................................... aTop of Active Fuel TSC ............................................................................................................. Technical Support Center WOG.............................................................................. Westinghouse Owners Group 201 Page APPENDIX B -DEFINITIONS The following definitions are taken from Title 10, Code of Federal Regulations, and related regulatory guidance documents. A4e4ALERT: Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels.G.ne.a. Efngen.e. GENERAL EMERGENCY: Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.Notificatieti all Un.usual EventUNUSUAL EVENT (-NOU-EIE4-U -0: Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.Site.Area Emi.e.gene..SITE AREA EMERGENCY: Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.The following are key terms necessary for overall understanding the NEI 99 01 emergency classification scheme.Emergency Aetien Level EMERGENCY ACTION LEVEL (EAL): A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given em'crgeney elassifieati"n le-'LEMERGENCY CLASSIFICATION LEVEL.Emergeny' Classifieati iaLel EMERGENCY CLASSIFICATION LEVEL (ECL): One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency elassificatieHn IevesEMERGENCY CLASSIFICATION LEVELS, in ascending order of severity, are:* .otificatiein of Uknulsial EventUNUSUAL EVENT (NOWE I.JE.SAiet4ALERT

• Site Area Emergency SITE AREA EMERGENCY (SAE)-General Emergency GENERAL EMERGENCY (GE).0 ......202 1 P a g e Fissioi Pr'oduct Barrier Th-esholdFISSION PRODUCT BARRIER THRESHOLD:

A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.Initiating INITIATING CONDITION (IC): An event or condition that aligns with the definition of one of the four emergency classification levelsEMERGENCY CLASSIFICATION LEVELS by virtue of the potential or actual effects or consequences. Selected terms used in Initiating Condition INITIATING CONDITION and EMERGENCY ACTION LEVEL Emergency Action Level EMERGENCY ACTION LEVEL statements are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below.CONFINEMENT BOUNDARY: (Inse,. a specific definition for this term.) Dc'vcloper Note The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage.CONTAINMENT CLOSURE: (Insel. a site specific definition fo this term:.) Devclopcr Note The procedurall d ..efined cond.itionS Or actions taken to secure containment (priimary or sec.ndar. y for BAIR) and its associated structur~es, systenis, and components, as a functional barrier to fissio pr.od ct .el.aS. under shUtdowN'll c ,nditions. Those actions necessary to place the RCB in the closed containment condition that provides at least one integral barrier to the release of radioactive material. Sufficient separation of the containment atmosphere from the outside environment is to be provided such that a barrier to the escape of radioactive material is reasonably expected to remain in place following a core melt accident.CREDIBLE SECURITY THREAT: Information received from a source determined to be reliable (e.g..law enforcement. government agency. etc.) or has been verified to be true or considered credible when: (I) Physical evidence supporting the threat exists, (2) Information independent friom the actual threat message exists that supports the threat, or (3) A specific known group or organization claims responsibility for the threat.EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present.FACILITY: The area and buildings within the PROTECTED AREA and the switchyard. FAULTED: The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized. Dcveloper-Note T.is ter .is applicable o only.203 1 P ag e FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.HATCH MONITOR: Temporary monitor installed when Containment High Range Radiation Monitors RT-8050 and RT-8051 are out of service.HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station.HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.NORMAL LEVELS: As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value.OWNER CONTROLLED AREA: (Insert a site spc.ifi. definition for- this term.) De'e!Opcr Note This-tr pically thaken to' mean the site purOprty ouned by, r othePwisO uCndeR the C-ntrol of- the licensee.in sOme eases, it may be appOpCiate frs r -a Iiersee to define a smallers aea with a perinet, erclser to the plant Protec-ted Ar-ea perifnetcr (e.g., a site with a large OCA wher-e some porioins of the boundary, may be a significant distance from; the Protected Ar-ea).. in these cases., develeper-s should consider-using the boundary definied by the Restricted Or Secured Ownier Controlled Area (ROCAiSOCA). The area and boundary slected for shemife Else muitst be eeonsistenit with the description of the sam~e area and boundar-y cont-ained in the Secur-ity Plan.-The area Surrounding thle PROTECTED AREA where SIP Nuclear Operating Company (STPNOC) reserves the right to restrict access. search personnel, and vehicles.PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.PROTECTIVE ACTION GUIDES (PAG): Environmental Protection Agency (EPA) guides for protective actions to safeguard against radiation exposure friom nuclear incidents. 204 P a g e PROTECTED AREA: (Inser.t a speefic ,definitin fer th.is term.) Dev.loper-Note This t.rm is typically taken to iean the The area under continuous access monitoring and control, and armed protection as described in the site Security Plan.REFUELING PATHWAY : (Insert a site specific defin.iti.n fc. this term.) Deeloper. Note This des.riptiOn ;incl,, Includes all the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.RUPTURE(D): The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection. DeVe!Oper Note This term is applicable to PWRS only.SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related. Dc'velopcr Notc This ter.m may be modified to include the attributes Of "safety5 related" in, accordancee with 10 CER 50.2 Or other Site specifi termliineoffg, if desired.SECURITY CONDITION: Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.SECURITY EVENT: Any incident representing an attempted, threatened, of actual breach of the security system of reduction of the operational effectiveness of that system. A security event can result in either a SECURITY CONDITION or HOSTILE ACTION.SITE BOUNDARY: The edge of the plant property whose access may be controlled by STPEGS. This boundary is congruent with the Exclusion Area Boundarv for the purpose of offsite dose assessment. UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally.UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.THYROID CDE: The dose equivalent to the thyroid firom an intake of radioactive material by an individual during the 50-year period following the intake.VALID: An indication, report or condition is considered to be VALID when it is verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. This may be accomplished through an instrument channel check.response on related or redundant indicators, or direct observation by plant personnel. The verification methods should be completed in a manner the supports timely emergency declaration. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. 205 1 P a e e APPENDIX C PERDACANENTLY DEFUELED STATION T/sEATLs Reeognitioni Category PD proevides a stand alon~e set 4f I4s/EA6s fat- a Per-manentfly Defuieled nuelear pov.'cr planit to consider-for- tis ini dcvelopinge a site spccific emergency, classificatio schemie. For- development, it was assumied that the plant had operatedd unM.d~er a 10 GCFR, § 5 0 lieenfis e and that the operating company has pe.a.nen.tly ceased plan. t oera..tionS. Further, the company intends to store the spent fu.el within the plant f.. r som. e period of tim.. e.When in a permaniently deflieled condition, the plant licensee typic-ally r-eceives approeval fl-em the NRC for- exemiptioni froem specific emer-gency platnin requ.ireents. These exNemptions r-eflect the lowered radiological, souce term and risks associated with spent fuel pool storag- e rtelative to r'eactor at powAer operationl. Scuree termiis and accident analyses associated with plausible accidents ar-e documented in the station's Fina.l Safety Analysis Repo.t (FSAR), as updated. As a result, each licenfsee will nieed to develop a site specific emergency classification schemie using the NR appo.e eem.ptions, revised source terms., and revised aeeident analyses as documented in the station's FSAR.Recognitioni Categtory PD uises the samfe EC~ns as oper-ating reactors; however, the source term and accident analyses typically limit the EChs to an Unusual Event and Alert. The Unusual Evenit W~s prov idefo an incrieased-awrns f abnormnal conditions while thie Alert l~s are specific to actual or potenitial impacts to spent fuel. The source teigms mnd release motive forces associated with a permanently defueled plant wouild not be suffieient to r-equlire declar-ation of a Site Area Emergencey or- Gfeneral Emerg~ency. A permaniently defueled stationl is essentially a spent fuel stor-age faeilil' with the spent fuel is stored in a pool of water that se..ves as both a cooling medium (i.e., removal of decay heat) and shield friom direct radiation. These primaaryj funtitions of the spentskiel storage pool ar-e thle focus of the Reognition Categor.y PD Ws and EALs. Radiological effluient , I anid EALs we"re included to poiea basis for- elassifying events that cannot be readily' classified based oni ani obseirvable events or planit conditions alone.Appr-opr-iate W~s and EAL-s from Recoagnition Categories A, C, F, H, and S wvere modifiedad included in Recognition Catego,')F PD) to address a spectrum of the evenits that may affect a spent fuel pool. The Recognition Category, PD j~s and EALs reflect the relevant guaidance ini Section 3 of this docuent. (e.g., the importance of avoiding both over classification and under- classifieatio:). Nonetheless, ea. h li.ensee w.ill need to devlop their- emergency classification seheme usig .the NRC approved exemptions, and th es oure te.. ..' anid aeeident analyses specific to the lieensee.Security related events will also need to bOe onsidered. 206 p aP e Table PD 1! Rceognition Category "PD" initiating Condition Matri jSUAL EVENT ALERT PD AUI Release of gaseous or liquid radioactivity glreater, than 2 times tile (S-ite speeific efflenit release controlling doetumient) limiits for- 60 minuites or- loniger.Op. 4Aides.- Aola pplibet PD AU2 UNPLANNED rise in plant r-adiationi levels.Op. oIdes o.. IApplki PD SU4 UNPLANNEDP spent fuel pool Op.. Modf.. : .... A pp÷ PD HII~ Confirmfed SECURITY COND ITION ar- threat.Op9. Ahodes: o Appeieable PD HU-2 Hazardouis evient aff-ecting SAFETY SYSTEM equipment necessary forspent fuiel cooling.Opý. Modes:- Not Appike&oblek PD 1MU3 Other" conditions exist which in the judgmlent of the Emergency Director wlaffaflt dleclaration of a (NO)UE.Op. Moeeks: AW~f/ Ipplicbl PD AMI Release of gaseous or liquid radiactiity esuling in k fsite do)se gr-eater-than 10 mr-em TEDE or 50 nirem thyteid -PDE Op. Modes." N: Applipcble PD AA2 UNPLANNED rise in plant r-adiation levels that impedes plant accessT ruired to maintain spent fitel integrity. O.Modes. No pphblet PD HAI HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne at4ack threat within 30 minutes.Op. Mogdes: Nor. 1pp!ieefbk-PD HA3 Other conditions exist whic-h in theEjudgment of the Emergency Director warrant eclelar'ation Of an- Alert-.Op. Modees: No XqApj5ie.Thk Table intended (Revis ion 6)Is P tr uise btv L/\L deveieoneis. inclusion in licensee documients is net required. NEI 99 0!i oNle I- 11IN111- i.. ....... l .............. ..PA i r 207 Page P)r ATi ECWL Of UUSuaIl initiating Condition: Release of gaseous or liquid radioactivity grcater than 2 times the (site speeific eft.ent r.elease .ontrolling do.ument) limits for 60 minu.tes or loger.Operating AModc-i Appli-bility: Applicabl-e Example Emergei.n.y Action Le:,ls, (I cr2)Notes;" The Emergency Dir. e.t. should d .lare the Unusual Event pr.mptly upon deterining. that 60 mintes has been ekeeede, ofr will likIely be eic-eeded." if an ongoinlg release is detected anid the release stait time is unknoewn, assume that the reles duration has exceeeded 60 minutes." if the effluent Pawv, past an effluent monitor is known to have stopped due to actionis to isolate the release path. then the effluent monitor- reading is no lonager- valid forF cla'ssification purFposes. (1) Reading on ANV effli....t radiation monitor. gr.eater. than 2 times the alarm setpoint established by a currfent radioactivity dischar-ge perm~it for 60 minuites or longer.(2) Sample anialysis for agaseous or liquid release indicates a oncentrIation or release rate greater than. 2 times the (site specific effluent release controlling document) limits for 60 minutes or IeingeF.This WC addresses a potential decr.ease in the level of safety ofithe plant as indicated by a low l.v.l r-adiological release that exceeds regulatory commitments forf an extenided per-iod off timle (e.g., an unconti-olled r-elease). it includes any gaseouls or- liquid radiological release, monitor-ed or- an monitor-ed, inc.luding9. those for which a radicaetivity discharge perm.it is n.rmall. ' pr.epared. Nuclear power plants incor-porate design feaatures intended to controal the release of radioactive effluents to the environment. Further, ther-e ar-e administrative controals established to prevenit unintenitionial releases, and to controEl anld monEnito-r intetpion~al r-eleases. The occGUrrenc-e of an extended, unconitrolled radioactive release to the environmient is indicative of degradation ini these features and/or conRtrols. Rad.ioogical effluent EAs al r also inclu.ded to provide a basis for. classif..ing even-ts an. d oanditi n that cannot be readily or appriopriatel y classified on the basis of plant conditions alone. The inc.lusion of both planit condition and radiological effluent EALS mor)e fully addresses the spectrum of possible accidenit eents and conditions. Classification based oni effluent monitor- readings assumes that a release path to the enivironfment is esabisihed if the effluent Dlow. n.t ani efftluent maniteir is known to have stonned due to actions to isolate thle release path, then the efflue ft mIn itoHR readiing is no lo..ger. valid for pu.pos..208 P a g e Releases should n1t be pr;rated or aver.aged. Fr ehamp,, a release 1ieeding 4 ti.iieS relcase lim;its fr 30 minutes does not meet the EAL..E1AL -.] This EAL address~es, radioaetivity releases that cauise effluent radiation nainitOr readings tO.... .2 times the limit established by a radioa.tivity diseharg" permit. This EAL will typically-be associated with planned bat.h r.eleases. fm... 110.. c.On.tinu.s release pathwa.s (g..radkaste waste gas).EAL #2 This EAL addr-esses uncontrolled gaseouis or liquid releases that are detected by samploe analyses or en-ir'nmental s-eys, particularl) on ..n.onitored pathways (e.g., spills of radioacti.ve liquids into storm drains, heat ex'changer 'l leakage in river- water. systems, etc.).Escalation of the emer..gency classification level would be via !C PD AAI.DeveloperNotes; The "site specific effluient release controllinig document" ais the Radiological Effluenit Technical Spec.i.fi.cations (RETS) Or, fo. plants that have implemented G.en.e.ic. Letter. 89 0.111, fte ffite Dase Calculation Manual (ODCM). These documenits implemient r-eguilationis related to effluent controls (e.g., 10 C.F.R Part 20 and 10 CFR Pait 50., Appendi-x .). As appr;opriate, the RETS or ODCM methodology should be used for establishing the Monitor thr.esholds for this !&C..Listed monitor-s should incelude the effluent monitors descr-ibed in the RETS or- ODCM4.Developers may also consider inceluding installed monitors associated with other potential effluent pathways that are noet described in the RýETS or ODC-N1213. if incluided. EAL- val uels fo-r these monitors should he determiiined usinig the most applicable do~se,/release limits pr-esented iii the RETS or ODCN'1. it is reogizdthalt -A clalcula-ted EAL value may be below v.hat the mon)IitorF caii read; in that ease, the monitor does not need to be included in the list. Also, some m~onitors may not be go','ri~ed by, Technical Specifications or- other license related r-elated requiirem~ents; ther-e4foe, it is ipratthlat the associated EAL and basis section clearly idenitify, any limitations on the use or availability of these monitors.Some sites may find it advantageous to address gaseous arid liquid releases wit separate E=AL.Radiation monitor rekadings, shoulýd reflect valuies that c-orrespond to a radioloigical release exc-eeding, 2 times a release controel limit. The controellinig documenit typically descr-ibes m~ethodologies for deter-mining effluent r-adiation mAonitor SetpEointS; thteae mlethoedoAlogies should be used to deter-mine EAL values. in eases wher-e a methodology is not adequately defined, developers should deter-mine values consistent with effluenit control re-gulations (e.g., 10 CFR Part 20 anid 10 CPR Part --O) Appendix I) and related jguidanc-e. 11 Inplmeozeqta.'ion of .D:.gog;na~tic Cew;tro~fiqr Radiologicaql W//hent Teehniea! Spec/eat icon~s i? tAe ldA4;inistra!ive C'on;iioe Sectiieof /the Technicial Spqecifications and, the Reloeentlo qof Procdual Detai4s of RE-TS to r a to ese H r eil 111W el rf9rt3 ieeeszy Wiffe i R=i dill 12 This incluides consider-ation of the effluent mionitors descr-ibed in the site emer-gency plani seetion(s) whic address the requirements "of 10 CFR 4 .47(b)(8) and (9).13 Dy .lp. s.hould keep in mind the requir.emen.ts of 10 " R 50.54,(q) and the guidance provided by INPO related to emergency response equipment when conisidering the addition of other effluent monitors..... ...................... 2.... ..........P...N 20 9 1P age For EA6 ft Valkies iii this RA L should be 2 times the setpoint establi shed by the radieaetivity discharge permIit to war-n of a release that is not in compliance with the specified limits. indexing, the value in this rnanncr enisures eansistcncey between the EA6 and the setpoinit established by a specific dischre permit D..... JEPeh ..OE.ld reSear.h r .adiation monitor design documents or other information souces to ensurfe that 1 ) the EAL value being considered is within the usable respon.se and display r.an.ge of the instrument, and -2) therc nrc no automatic featur-es that miay render the moinitor r-eadinig inivalid (e.g., ani auto purg,,e feature triggered at a par.ticu.lar indication level).it is r-ecognizied that the conidition descr-ibed by this WC may resuilt in a radiological effluent value beyond the operating or display .an.ge of the installed effluent monito.. in those cases, EAL values should be determiined with a margini sufficient to enisure that an accurate moniitor-reading is available. For example, an EAL r reading might be set at 90% to 959% afthe highest ac...ate monitor reading. This provision notwithstanding, if'the estimated/calculated monitor" reading is greater than" ar ... iately 1 100A of the highest accur.ate mo.itor reading, then developers may choose not to include the monEiitorF as an indication and identi6,' an alternfate EAL threshold. Inidications froEm a rcal timne dose projection system are not included in the generic EALs. Many licensees do noat have this capability. For those that do, the capability may not be within the scope of tho pln Teenneai Specifiations. A li-enisee mnay request to include an PA/\L using real time dose projecto sytem results; approval will be considered on a case by ease basis.Inidicationfs from a perimeter-monitoring system are net included in the generic EAI~s. Many licensees noet have this capability. For- those that do, these molnitors may not be controlled and maintained to thqe same level as .lan. e.ui.men .. o. within the s.... of the tlait. Tech'nical S ,ecimfiations. in addi.io., do r;eadings may be influenced by environmental ...ther factrs. A licensee may request to ii usn F er-imeter MOnitoring systemf; approval w.ill be considered on a case by case basis.ielude an EALýECL Assignment Attr-ibu,-tes .l N"D I 99 01 (Revision 6)1 III~II I ý ---, ý -') f) I ') C'-,&2101 Pa g e ECL: Notification of Unusual Event initiating Condition: UNPLANNED rise in plant rAdiAtion; Ilev~l.Op.r..ting Modec Applicbiliot.. Not Applicable Example Emcr-gcncy Action Lcvcls: (I or- 2)(1) a. UNPL..NNEDwater. level droep iii the spenit fuel pe.l as indicated bYANY fthe fbllowing: (site speeific lev-el inidications). AND b. UNPLANNED ise in area Fradiation levels as indiated by ANY Of tie fOllwINlg radiation mn:ti4e!.(site specific list of a.ea radiation monitors). (2) Area r-adiation monitor reading, or surv'ey result indicates an UCNPLA.ThED rise of 25 mR'm- over NORMAL LEVELS.This IC. addresses elevated plant r-adiation levels caused by a decrease in water level above irmadiated (spent) fuIel Or ther"F LD events. The ..a.. q ra tio;n --.levels ae indicative of a minor loss in the ability to conitrol radiation levels within thie plant or- radioactive materials. Either condition is a potential degradation in; the level of safety of the plant-... ~ le.eldecre.ase Will be primar.ily. deter.mined by indi.ations fro.. .available level inst..mintatior .Other: sourcees ef leve'l indicationis may, incelude repeits frmplait per-sonnel or video camera observatiefns (if available). A sigxnifican~t droep in the watrlve may, also cause anicraeinlte radiation levels Et adjacent areas that can be deiected by moitrsin ~ tho eloatEions. The eff.ets of planned ev'olutions so-'here the elevated reading is due to rea sthat result from plaaned a radioac-tive w.aste materials. ot a ct;ild be considered. Note that EAL #1 is applicable only in cases i U1NPLANNED water lcv~el drop. EAL #2 exceludes radiation level vities suchi as use of radiogr-aphic sources and movemento Escalation otthe enicr.ency:' classitication level would be via IC PD AA! or PD AA2.2111 P a g e Dc cIopcr Notcs.2 Far- EAL #I Site specific indieationis may incelude inistr-umentation values such as .vater- level and area r-adiatiOn H101140F readings, and porsonnel i-epei~s. 4f available, video eameras may allow. fair remote observation. Depending on available instrumnentaticn. the declaration may alse be based on indications Et water makeup rate and/cr decre-ase-s in the level of a water sicrage tank.For- EAL 42 The speeified value of 25 mRlhi- may be set tEo another- valuie for a specific applicationi with appropriate justification. ECL Assignment Attributes: 3.1. 1.B 212 1P a g e P~D sui ECL: l~otifieation of Unusual Event lnitiating COndItion: U N PL0A N VSpent !Uci pool1 te~perature rise.m m 1 AT .A ,Cfi-amw M~'odeC Aoinflicmltv: [Not Aesniteigaie le Emcr-gcncy Action Levels-.U[NPLA[N[NE sn~ent +mie seal temioeraturce riSe tO freater thani Wste snceetme F--I BftsS.This WC addr-esses a eondition that is a pr-ewusor-to a mresi event and repr.esents a potentia degradation ini the level of satety of the plant. Wt uncorrected, tollin flý- tile P88 v1 Wiloc-cur!, anid resulit inia loss of pool- level and- incr-eased radiationi levels.Escalation of the emergency classifiation level would be via IC; PD) AA 1 or PD) AA2.Developer Notes: Tha site specific temperature should be chosen based on the stalling point for fuel damage alclationsin the SAR. Typicaly,ý this temfper-ature is 125' to I 50' F. Spent Fuel Pool temaperature is nonmillyv maintained well below this point thus allowing time to correct the cooling system mlueinpiit classific-ationi. ECL Assignment AttribUtes: 3.1 .1 .A N-E1 99 01 (Reiisi F) n 6)ýSIAI '"Q l pI.~ 10 11 C2 Q 213 Pati e PD BUI VVI NcWA4tPP16R Ain If 11411qiil ýVPRnt iniftiti ig JConclitn! Confirmie SELUK1 LY COUNDITIION op 4hreat.Opcrating Mode Applicability: Not Applieable E ample Emer-geney Action Levels: (1 or- 2 orý 3)(1) A SECURITY CONDITION that d... not involve a HOSTIL6EACTION as r.epoted by the (site specific security shift supervision). (2) Notlaifiiation of a credible security threat direted at the Csite (g) A validated notifi.ati. n fr... thle NRC .rvd .ro of an air0Caf 1 .hreat.gftsi5+.This lC addresses evcnitc that pose a throat to plant perconinel or- the equipmient nececcary to miainitain cooling, of spent ftel, and thus r-epresent a potential degr-adation in the level of plant safety. Security events whichl do not mieet one of these EALc awe adequately addressed by the rcquireienitc of 10 CFR §73.71 Or 10 CFR § 50h72. Security events assessed as HOSTILE ACTIONS are clcifa leinder IC PD Timely anid accurate eommnunieatiainc between SceurK' Shift Supervicion and the Controal Roomn-i fol proper classification Of Acu , related event. of these eveiltc will in"itiate aporate threat related notifiaticoc to plant personnel and OR-(c.Secuity plans and ter.minolo.. .a.e based on the guidan.e provided by NEI 03 12, , Stecurih-Plan?, Trainii;hg an~d Qial~ieiationi Pkfm. Saf -egArd otne Plan [Hid hulopendent Spent Fien Storage hncta!!LAtief Seclurity A-agranz]. EAL 41 references (cite specific security shift super-vision) because these are thle individuals trainied to confirmff that a cceuriitv event is occurring Or has occurr~ed. Tr-ainincg oni security event confirimation and classification is controalled due to the nature of Saf~elguar-ds and 10 CFR § 2.39 information. EAL 42 addresses tlh receipt of a .i. dible security threat. The c-rediblity of the th.eat is assesscd in accor-dance with (cite specific proceLdure)-. EAL #3 ad-d-ressesr thie thratrom the- impact of an aiircraf on the plant. The NRC Headquarteirs Operations Officeri (HOG0) will commiuniceate to the licensee if thie thr-eat involves ani aircrfaft. The status A .... .... ..... I ... ..4 X n l 1 1 1 XIDI ... ....... .. ... .. .. .... .... .. ...4 :D .... .... s u performied in accordanceewith (site c pe iimiibim 1tie Li,. a a OR E)i~iiii tL Lii tiii, eat s Emergency plans and implementing procedures are publi. therefo.re, EALSs hould no."t Incrprte Security sensitive inforimationi. This incluides informfationf that mfay' be advantageouis to a 214 P a ge p cteniial ad'vcrzary, sueh as thie panie tilafs eoneerning a speeifie tlifeat of threat lacation. Sec a ed ill nOn p4bl6 doc-HfentS SuH~l aS tile Securit)' Planl.t":ty-sefis t N e n i mat Oll S !E)U c P-tvilta 2151 Pa ae Escalation of the em..ergen.y .lassitcati.n level would be via 1C PP H,,AI.DeveloperNtes: The (site specific sece wiity sihift supervision) is the title of the en shift inoivicluai ressenstioe ioil supfer-VISIon -T ...- onl bnll SeEHFIENY for~e.The (site specific procedure) is the proceduire(s) used by Control Roomf and/or Secur-ity personnel to deter-mine if a secuirity threat is cr-edible, and to validat-eeito aircr~aft thfeat information. Emergency plans and implementing pi-oceduite are public documents; therefore, EAbs shouldnt inoprte Security sensitive infor-mation. This includes informnation that mfay be advantageous to a ptential adversa,', su.h as the pa.iculars aonfcerning a specific ti.reat or threat location-. Seceurity1 sensitive inform.ation should be conItained in non public docum.ents suehl as the Sce....ity Plan.With due consideration given to the above developer note. EAbs may contain alpha or numbered r-eferences t s elected events descr-ibed in the Security Plan and associated implementin .......es Such references should not contain a recognizable description of the e e F.o e ,aple an EAL may be....ded as "Secu.ity event #2, 45 Or #9 is r.eported by the (site speci..c. secuity shi.. supe rvisioen). EC-L Asso ig~nient Attr-ibutes:

33. 1.1 .A NEI 99 01 (Re vosinn 6) Nm--rem:htar 201 CI IGI 2161 Page PDIIU)VC-1 N- IItif4P~tifjr,,'l A4: 11n1Ii'il rF.'"Pt 1nitrntin~LonditMan p;arou eavent atteetmo jig E Q '%u Y Q- Y'.[M equipment neeess.a~.

ta-r sp~ent A iel-7 -................... j .... [ .......... e6811g Oprniting Mode Applienbility: Not Applicable le Emergency A.tion Lcvci...The occurr-ence of ANY of the fellowi" Seismic event (earthiquake)" Internal or external flooding event" High winds or tornado strike EXPLOSION*(site specifie hazar-ds)*.Other events with similar- hazarde ng nazaroonis events: flaraeter'isties as cleteizninedl hN the Shi# l anacer-AND~b. The event has damfaged at least one train of a SAFETY SYSTEM needed for spent fuiel coling, e. The damaged SAFETY3 SYSTEM tr-ain(s) eannoet, or potentially cannot, function based on EITHER: 0 Indicationis of degraded per-formance 0 VISIBLE DAMAGE perfform its de in This lC addr.esses a hazardo..s event that .auses damage to at least one tr.ain of a SAFETY SYSTEM needed fer spenit futel coolinig. The damage muibst be of suifficient miagnitude that thie systemi(s) tr-aini cannot, Or potentially cannot, perform its, design function. This condition reduces the margin toa less or potenitial less of the fuiel clad barriOer. amid therefore r-epresents a potential degr-adatioii of the lev.el of safety For EAL Ie ', the Pfist bullet addresses damage to a SAFETYV SYSTEM train that is ini service/oper-ation since indications foi- it will be readily available. For EAL I., the seiond bullet addresses damage to a SAFETY SYSTEM train that is not in r seR.ice/oeroti oradily appar-ent tkletigli inidications alonie. Operators will1 make this deteffiniation basd n he totality of avail-able, event and dlamagge r-epor-t infor-mation. This is intended to be abie assessmient not reqluiring lengthy analysis orF quantification of the damage.217 P a 1 e 6EsalatiOni Of the emfergency classification level eould, depending upon the event, be based an any of the Alert W.S; PD AA , PD AA2.. PD HAl.. r. , of P1A3 Devlper-Ntes: For (site specific hazards), dev~elepers should c-onsider inclueding other significant, site specific haz6ardS tO theQ bulleted list contained in EAL !.a (e.g., a seichie)-. Nuclear power plant SAFETY SYSTEMS are coznprised of two or more separate an.d redundant trains of quipmnent, in accordance with site specifie design criteria.EC-L Assignment Attributes: 3.1.1 A anid 3. 1 1C 218 Page PPT T-T Ff1 ~J~tific~th~n "f' Jnwmnl F"ent initiating Condition: Otheir eonditionlS exiSt WhiCh ini !he jUdgmienit Of tile E rgil0,ency Dir1ectorI WarrFant declaraton of a (NO)UE.Operating Made Applicability! N]ot Applicable Ex.ample Emcr-gcncy Action Levels:i (1) O~th~e conlditiEnS eXiSt which inl the jubdgment of the Emergency Director indicate that events are in rogessorhave oeeurrced whichi indicate a potential degradation of the level Of Safety Of the p~lant Or indicate a security threat to facility pro~tectiOn has been initiated. No r-eleases of radioac-tive mateia reurnaff-site r-esponse or monitor-ing are expecte-d unless furFther degrFadatieon of safety systems occurs.This 4C addr-esses uinanticipated coniditions noat addressed cxiplieitly elsewhiere butt that warranit deelar-ation o-f an emfergency because conditions exist whichi ar-e believed by the Emer-gency Dii retor to fall under thle eimer-gency classification level descr-iption for a NOUE.219 1Pa e PD A1 ECL: .4e44 initiating Condition: Release Of gaseouS OF liquid radioactivitY resulting in 6ffsite do)Se greater- than 10 mrnem~ TEDE of 50 mrcm thyroid GCDE Operating Mode Applicability: Not Applicable Example Erncrgcncy Action Lcvcls: (I 1 r-2 or 3 or '-z1)Notes: 0 The Emer-geney Director-should declare the'Alet4 promptly uponl detcr-mining, that the applicable time has been .......... will likely be e..eeded.if an ongoing release is detected and tle relea start time is unkow.. n., assume that the.. rele.ase duraion has exceeeded 15 m~inutes.If the effluent flow, past an effluenit monitor- is known to have stopped due to actionis to isolate the release path, then the ef-flent monitorir .eading is no Valid for classification The pre calculated effluent monitor- values presented in EAL 41 should be used for- emqergency classification assessments until the results fro.. m a dose assessm.ent using actual meteoerlogy are available.(H) Reading oni ANYý of the foljlowing r-adiation moitor48s greater thani the reading- shownA, forF 15 Sminutes or longer: (site specific monitor list and threshold values)(2) Dose assessment. ui actual. eteorology indi.ates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or bey.on.d (site specific dose ie .. pt... paint).(3) Analysis Of a liquid effluent sample indicates a concentration Or r~elease rate that would result in doses greater thanl 10 ireina TEDE or- 50 mrem thyroid CDE at or beyond (site specific dose receptor point) for one hour- of exposure.(4) Field surve' r@esiI~ts indic-ate EITHER of the following at or be:y'oid (site specific dose receptor" Closed window. dose rates greatei- than 10 mR'hr ex.)pected to continue for 60 minilutes Or" ~Atfalyses of field survey samples indicate thyroid CDE; greater- thani 50 mrem for one hour' Of i llalaion Bafts This WC addie. e a release of gaseous oe liquid radieativit' that results in projected or actual offsite dese gfeater than or equial to 11% of thie EPA Prot!ectilve ActiOni GUides, (PAGS). it includes bot 220 1 P a tz e moniitored and tin moniitor-ed r-eleases. Releases of this nugniitude represenit ani actuial or potenitial-s-ubstaintial degradation of the level of safety of the plant as inidicated by a 2211 P a ge radiOlI8-cal release that signifi.antly excees r.g.latoi.y limits (e.g9., a signifiant release).Radiologxical effluent EALs are also included to prov~ide a basis Bfr elasský'ing evenits and conditions that canno! be r-eadily of appropr-iately, classified on the basis of plant conditions alone. The inclusion of boeth plant condition and r-adiologgical effluent EALs mor~te fullyý addres-ses the spectrumff of possible accident@eventS and coniditionS. The TEDE dose is set at 10% of the EPA PAG of 1,000 mr....m while the 50 mrem thyroid C-DE was established in consideration of the 1:5 ratio of the EPA PAG foir TEDE and thyroid CDE.Cl 1a ss itication based Em effluent meontor r-eadings assumes that a release path to thc environment i established. if the effluent flow past an effluent m.nitor is ..own to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longe.r valid fre ElassifiEation purposes.Devlper-Noes.! While this IC- may not be met absent chiallenges to the cooling of spent fuel, it provides classification diver-sity and m~a:, be used to classify ev~ents that wouild not r-eachi the same ECL6 based on plant condlitions The ERA PAGs; are expressed ill term-s; of the sum of the effective dose equivalent (E=DE) and ýthe committed effective dose equvalent (CEDE), or as the thyroid committed dose equivalen"t (CD9). For the purpose of these CEALs. the dose quantity total eff-ective dose equivalent (TEDE), as defned in 10 CFR§ 20. is used in lieu of".. .sum of EDE and CED.. 1 The EPA PAG uidan fe provides fora the use adult thyroid dose cofnversione freeaose ,sot :nix;me states have decided to base protective actions on child thyroid CDE. Nuclear power plant !Gs/EFALs need to be consisitent With thle pro~tective action methodologies employed by the States within their EPZs. h thyroid CDE dose used in the 1C and EA~s should be adjusted as necessary to align With State protectiv~e ac-tion decision making riera The "site spec-ifi monitor list and thr-eshold values" should be determlined With c-onsideration of thle 0 Selection of the appro~priate installed gaseous and liquid effluent monitors." The- effluent moEnitor reading-s should corrfespond to A- do-se. of 10 nr-em TEDE or 50 mrem thyroid CDE at the "site specific dose r-eceptor point" (consistent with the calculationl miethodolog employed) for one hourj of exposure." Monitor- readings will be calculated u~sing a set of assumed mcteorolegieal data or- atmolspherici dispersion factors; thie data Or factorFs selected for as@ should be the same -As tho A-s-e employed to ealculate the monitor readings for 1c; PDAU." The calculation of monitor readings will also) require use of an assumed reles iopic ix; the selected mnix should be the same as that employed to calculate monitor r-eadings for ic- PD AW1." Depending upon the methodology used to calculate the EAL values, there may be overlap of some x-hm* it n I S N.J. ps- i-) V 111111 '.I.1/4 ptAI vpnqL +..I3~fl ýIJ -v~lp7 1,J, IIofr;*I it -vxrtrra* v p Y 51 iiii ei4smeq A -E)e ea esea rtt an n t ea G 6.222 1 P a g e The "site specific dose Feceptor-peint" is the distance(s) anid/ori locatiens used by the lieensee to distinguish between on site and .ff.si.e doses. The sele.ted distance(s) and/r locations should ,elect the content of the emergency plani, and the proccdui-al methodology used to determine off-site doses and P~rotective Action Recommendations. The variation in selected dose receptor points mfeans there may be some differences in the distance firom the release point to the calcuilated dose point from site to site.Developers Should r~esearch radiationl mROnito deSign documfenlts or other infB~formatio sourees to ensurie that 1 ) the EAL value being conisider-ed is within the uisable r-esponse and display range of the instrumient, and 2) ther-e ar-e no automatic featur-es that may render, the monitor- r-eading invalid (e.g., an auto purge feature triggered at a pa~iciulair inidication level).it is r-ecognized that the conidition descr-ibed by this WC miay resuilt in a r-adiological effluenit valuie beyonld the operating or display range of.the installed effluent monitor.. in those cases, AL values should be determiined with a miargin sufficient to ensur~e that an accurfate moniitor reading is av~ailable. For example, ani FAL mionitor r-eading9 mioght be set at 90%O~ to 95% of the highest accurate monitor reading. Ti provi. si notwithstanding. if the estimated/calc-ulated monitor- reading is geate- tlhan approximately 110%11 of the highest accurate moniitor readinig, then developers may choose noet to includc the monitor as an indication and identify, an alternate EAL threshold. Although the W references TEDE.. field survey results ae generally av.ailable on.ly as a "whole body" do)se rate. For this reason, the field sur~vey EAL specifies a "closed window" survey reading.ndic.ations fro a real tim.e dose projection system are not included in the generFic ALs. Many li.ensees do not have this capability. For those that do, thie eapability, may not be w.ithin the scope of the plant Techniciial Specifications. A licenlsee mlay, request to incelude an EAL usinig r-eal timae dose prt-Eeetiom sytem r-esults; approval will be considered on a case by case basi's.Indications from a perimetcr monefitorfing, systemi arc noat inceluded in the generic EALs. Many licensees do not have this capability. For those that do, these mo m t be entrolled and maintained to the same level as plant equlipment, or- within the scope of the plant Technicial Specifications. hin additioni, readin-gs may, be influienced by enivironimental or- other factor-s. A licensee may request to incluide an EAL using a per-imeter molnitorinig system; approaval will be consider-ed an a case by ease basis.ECLI Assi..nmemfi Attrjblmtes: 3.1 .2.C-223 1 P a g e PD AA2 maintain spent fitil:m UNPLANNhD rise in plant radiation level that fte fieb -H4y-imeec-eas Biant access FequIreA Opernting Mode Applincbility: Not Applirabl e Exam~ple Emereenev Action Levecls: HI or 2)1) UNPLANNED dose rate greater than 15 mR'hr in ANY of the fo!!lowin.aesreurn contn~ios ocupancy to maintain; controlI of radicactiye maiteial or operation of systenms needed toa maintain spent fudel integrity: (2) UNPLANNED Area Radiation Monitor readinigs or survey r-esullts inidicate a rise by 100 mnRhr over NORMAL LEVELS that im~pedes access to ANY of the fellowkingf areas needed to mitiof radioactive material or" .peration. of systems needed to m.aintain spent fuel in.tegr.ity.(site specific area list)This WC addr-esses incrfeasied radiation levels that impede nlecessar-y access to ar-eas containiing equipmenit that must be operated manually or thiat requires local monitoring, in order to miaintain systems needed to maintain spent fuel integrity. As used here, 'impede' inceludes hinder-ing or initerfer-ing, provided that the in.terference or- delay is sufficient to sign.ificantly threaten n-ecessary plant access. it is this impaired access that re.su..lts in the actual o. potential substantial deadation of.the level of safety ofthe plant.This WC does not apply to anticipated tempoary-5 incrieases duie to planned events.Dcvelopcr-Notes: The value of I -i5 mR./h.r is der.ived from the GDC 19 value of 5 r.em in 30 days with adjustment for expected occupancy times. Although Section Il-.D.3 of NUREG 0737, .. #'T],fl.tion .PlaI........... , ipovides that the 14 m.R/hr. value can be aver.aged ov.er the 30 days, the value is u.sed ere without aeaigasa 30 day duration implies an event potentially more significant than an Alert.Thke specified v0alue of 100 mR/hr may be set to another value qfo a spoecifc applic-ation w.ith appr-opriate jati~eatien ECL Assignment ,Ar-iW;p., 4' 1 1 ?C 224 1Page PD 14A!nitiating ConEditior: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.Opcatig Alde Applicability: Nat Applieable f~amD~ nic mr-ffncv Action Lcveis:! OF 0r)(1) A HOSTILE ACTION is o "e'"ring or has .e.ur.ed within the .. ..ER CONTROLLED AREA as r.eported by the (site spe.ifi. sec..rity shift supervision) (2) A validated notiicationr from NRC of an aircrfaft a"nte: theat within 30 miWuEtes of the Site.Ba' Cndes ~ tecllefe faHSIL CINwthnteONRCOTOLDAE or notification of an aircraft attack thrfeat. This event will r-equlire r-apid response and assistance due to the pessipility or mne attacKi progriessing to the PRTPkU 2LL LUAKLA, or The need to prepar-e the plant and star for a potential aircraft impact.Timfely and accurate communici atiOnHS betweenl Security Shift Superv'ision anld the ConRol1 RcOO esential for proper classificatiein of a securfity related event.Security plans and termiinology are based on. the- guidance provided by NEI 03 12, Tcinplatofor 11 Ft'e!~g e'.ag !nsaaltio Securit hragraini. As tinme- and conditions allow, these events r-equirie a heightened state o~f readiniess by the plant staffand implemienitation of ensite protective measur-es (e.g., evacuation, dispersal or- shelter-ing). The Alr deelai'ation will also heighten the awareness of Offsite Response Or-ganizations, all owitig, them to be better- prepared shoulld it be necessary5 to con-lsid-er furt~her actions.This WC does not aenlv to ineidents that are ac-cidental events. acts of civil disobedience, or otherw~ise are not a HOSTILE ACTION per'petr'ated by a FORCE. Examples include the crash Of a small airera.., shots Trm .unte.. pnysia ........ , .......... employees. etc. KepOetlig 01 inese types ol events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.EAL fi1 is applicable for- any HOSTILE ACTION occurr-ing. or that has eocured, in the OWNE CONTROLLED AREA. This includes any action directed against an !SFSI that is located within the OWANER CONT-ROLLED AREA.EAL #2 addr-esses the th..eat from the im1-pact of an airciraf on the plant, and the anticipated arrival time is wihn30 minutes. The intent of this EAL is to enisure. tha thr -.ea reated notifiationis are made in.a timiely manner- se that plant personnel and ORC~s are in a heightene 2251 P ag e state Of readiness. This EAL is met when the throat :'elated information has b'-on validated in accordance with (site speeific procedure). The_ NýRC Headquarter-s Oper-ations Offeice (HOO) will comamunieate to the licepnse-e if thea threat involves an aifrcraft. The statuis and siz~e of the plane may be pr-evided by N0%R 4 lD through the NRC.in some eases, it may niat be readily apparent if ani aircraft impact with~in the OWNER CONTROLLED ARE was intentional (i.e., a HOST-ILE ACTION). it is expected, although net certain, that notification by an appropriate Federal ageney to the site w-ould clarfy this this case, the .ppropriate federal aeyisintended to be NORAD, F-B!, FAA or NRC. The emerfgency dleclaration, inceluding One based on other !Cs/EALs., should not be unduily delayed while awaiting notification by a Federal agency, Emer-gency plans and implementing procedures are public docum~ents; thier9efoe EALs should not inoprte Security senisitiv~e information. This incluides information that may be ad.anitaggeous to a ptniladversary, such as the paticueblars concer~ningg a specific threat or threat oain.Scrt sniieinformation should be contained in nion public documaents suchi as the Security Plani.Developer: Notes: The (site specific securityi) shift supervision) is the title of the on shift individual responisible for sup ersin of the on shift security, foree.Emergency plans and implementing pro.edu.es are public documents; therefore. .EA. s shold not incrpoateSecurity' sensitveifomtin This- inclujdes inform14ation that may be advantageouis to a-ptniladversar-y, suceh as the pafticulars concermnin a pcfi hea rthetloain.Scrity sensitive ifraonshou.-ld b6e cwontafinled in non public documents such as the Secur-ity Plan.Wioth due conisiderationeiigven to the abo'.e developer-note, EALs may conitain alpha or numbered r-efer-ences to selected events descr-ibed in the Securit-y Plan and associated implementing procedurFes. Such refei-ences should not contain a recognizable description of the event. Forf example, an EAL may be NNIrded aPs "Seetrit" ev~ent #2, 45 or- 49 is reporFted by the (site specific s~eurity shift super-vision)." See the related Developer: Note in Appendixi B, Definitions, for guidance on th~e development of a scheme definition for the OWNER CONTROLLED AREA.ECL= AssignmentI 0 226 1 P a a e PD 4A3 initiating Condition! Othe.i- cnditions exist whi.. h in the judgment efthc Emergency Dil'ector warrant declar-ation of aniAlert.Operating Mode Applicability: Not Applieable Example Emecgcncy Acti.n L.v.ls:.0) Other ..ndition.s exist whichi in thejudgment of the E."ergency Director indicate that events are inprgrssorhav~e occurred A~hiek involve an actual or potential sulbstantial djegr:adationH Of tile level of safety of the planit Or a Security' event that invle pi-ebable lif~e threatenling risk to site personnel or- dlamage to site equipment because of HOSTILE ACTION. Any r-eleases are expected to be limited to small fractialns of the EPA Protective Action Guideline exposurfe lev~els.This IC addresses unanticipated coniditions-not addressed explicitly elsewhere but that walfaiit declaration of an emergency becauise conditions eist hic-h -Are believed by the Emaergency Director-to fall under the eemer-gency c-lassification level descr-iption for an Alert.227 1 P a P e Attachment 4 STPEGS Emergency Action Level Deviation, Difference and Justification Matrix -revisions only STPNOC STPEGS Emergency Action Level Deviation, Difference and Justification Matrix Rev. 0 NEI 99-01 Rev. 6 Implementation APRIO4 STPEGS EAL DEVIATION/DIFFERENCE/JUSTIFICATION MATRIX TABLE OF CONTENTS FISSIO N PRO D U CT BA RRIER ICS/EA LS ......................................................................................................... 1 STPEGS EAL DEVIATION/DIFFERENCE/JUSTIFICATION MATRIX FISSION PRODUCT BARRIER ICS/EALS The following section is configured in a manner that is different from the Fission Product Barrier Tables in the STPEGS EAL Technical Bases Document. Where the Technical Bases Document evaluates all three fission product barriers simultaneously for a specific sub-category, this matrix evaluates each fission product barrier individually for all sub-categories. The significance of this fact is that where the fission product barrier table in the Technical Bases Document moves vertically through the sub-categories, this matrix moves horizontally. STP EAL DEVIATION/DIFFERENCE/J USTIFICATION MATRIX I Th rPchniri fnr i nA,, nr POTFNITIA1 I nAz jif ,rtaI tr A. Containment radiation monitor reading greater than (site-specific value).IMUL tippliLdUle H .Kt Kao iviomior RT-8050 or RT-8051 greater than 40 R/hr OR 2. HATCH MONITOR greater than 90 mR/hr 10ot Applicaole uitterence LOSS A.- see Ulobal Lomment#9 Loss A.1 -Calc STPNOC013-004 Rev. 2 lowered the setpoint.Loss A.2 -Lowered Loss A.1 resulted in lower Loss A.2 OR OR B. (Site-specific indications that reactor coolant activity is greater than 300 piCi/gm dose equivalent 1-131).3. Sample analysis indicates that reactor coolant activity is greater than 300 pCi/gm dose equivalent 1-131.Difference Loss B.- See Global Comment#9 C __________________________ U _________________________ i ____________ I ____________________________ Th.of RCS Barrier A. Lontainment radiation monitor reading greater than (site-specific value).Not Applicable Not Applicable Not Aoolicable Deviation Calculation STPNOC013-004 revised February 2015.Reliable radiation monitor readings not available due to the effect of Temperature Induced Current (TIC) and a setpoint value in close proximity to the background radiation level. STP EAL DEVIATION/D IFFERENCE/JUSTIFICATION MATRIX PnM Oqq nf Cnnta~inmant R2rrigar I nlr I-ntlplltii I Anc Not Applicable A. Containment radiation monitor reading greater than (site-specific value).Not Applicable A 1.RCB Rad Monitor RT-8050 or RT-8051 greater than 380 R/hr OR 2. HATCH MONITOR greater than 840 mR/hr V i 1I11U A.1 Calc STPNOC013-004 Rev. 2 lowered the setpoint.A.2 Lowered Potential Loss A.1 resulted in lowered Potential Loss A.2_______________________ -I- ________________________ 4 ______________________ __________ .L _________________________ Attachment 5 Supporting documents CALC. NO. STPNOC013-CALC-002 jEN ER Co N CALCULATION COVER SHEET REV. 2 EN, Ieodfy.PAGE NO. I of 47 Title: Radiological Release Thresholds for Emergency Client: South Texas Project Action Levels Project: STPNOC013 Item Cover Sheet Items I Does this calculation contain any open assumptions that require confirmation? (if YES, Identify the assumptions) 2 Does this calculation serve as an "Alternate Calculation"? (If YES, Identify the design verified calculation.) Design Verified Calculation No.3 Does this calculation Supersede an existing Calculation? (If YES, identify the superseded calculation.) Superseded Calculation No.Scope of Revision: Revision I incorporates decay time of one hour from shutdown as well as migration into Attachment

1. Change statement of no decay in the STAMPEDE runs.Revision 2 resolves reversed half-lives between Xe-133 and Xe-133m, Xe-135 and Xe-135m, and Kr-85 and Kr-85m in Attachment

1. Changes made in Revision 1 are propagated through all final results tables. Minor grammatical changes.Revision Impact on Results: For Revision 1, values calculated in Attachment I decreased and have become the limiting values.For Revision 2, main steam line monitor reading thresholds increased friom 3.90 to 4.03 pCi/cc for an Alert. Site Area Emergency and General Emergency thresholds increased from 39.0 to 40.3 and 390 to 403 ptCi/cc, respectively.

Study Calculation 1 Final Calculation Safety-Related[ Non-Safety Related[(Print Name and Sign)Originator: Caleb Trainor .- Date: 2/17/2015 Reviewer: Chad Cramer Date: 2/17/2015 Approver: Date: 2/17/2015 Marvin Morris CALC. NO. STPNOCO03-CALC-002 tE N E C O N CALCULATION Everyday REVISION STATUS SHEET REV. 2 PAGE NO. 2 of 47 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 0 03/03/2014 Initial Issue 1 3/14/2014 Resolve inconsistency in decay times for the two calculations. 2 2/17/2015 Changes to half-lives in Attachment I PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION 1-9 2 ATTACHMENT REVISION STATUS ATTACHMENT NO. PAGE NO. REVISION NO. ATTACHMENT NO. PAGE NO. REVISION NO.I 10-22 2 2 23-29 2 3 30-47 2 E N E R C 0 N Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 for Emergency Action Levels REX'. 2 ExcePpen e-Every project. Every doy PAGE NO. 3 of47 Table of Contents 1.0 O BJECTIVE/SCO PE .............................................................................................................................. 4 2.0 SUM M ARY O F RESULTS ..................................................................................................................... 4 3.0 M ETH O D O F ANA LYSIS ...................................................................................................................... 5 4.0 INPUTS .................................................................................................................................................... 5 5.0 REFER ENC ES ........................................................................................................................................ 5 6.0 ASSUM PTIO NS ..................................................................................................................................... 6 7.0 STAM PEDE CALC ULATIO NS .................................................................................................... 7 7.1 Unusual Event -RUI ................................................................................................................................... 7 7.2 Alert, Site Area and General Emergencies -R4 I, RSI, RG I ................................................................ 8 Attachm ent 1 -Hand Calculations .................................................................................................................... 10 Attachm ent 2 -Calculations .............................................................................................................................. 23 Attachm ent 3 -STAM PEDE O UTPUT ............................................................................................................. 30 CALC. NO. STPNOC013-CA LC-002 E N E RC0 N Radiological Release Thresholds REV. 2 for Emergency Action Levels REV. 2 Excel en e.--,Evety poit Ee',,dy"a PAGE NO. 4 of 47 1.0 OBJECTIVE/SCOPE The purpose of this calculation is to determine the Emergency Action Level (EAL) threshold values of a radiological release from the Unit Vent or Main Steam Lines for an Unusual Event, Alert, Site Area Emergency, or General Emergency. The calculated threshold values are to be included in the STP EAL Technical Basis document, which implements the new NEI 99-01, Revision 6, Emergency Action Level Scheme and will be submitted to the NRC for approval. Upon NRC approval, the values will be used in OERPO1-ZV-IN01, Revision 10, Emergency Classification. Both a hand calculation and the South Texas Assessment Model Projecting Emergency Dose Evaluation (STAMPEDE) software program were used to generate the results. The hand calculation is included as Attachment 1.Revision 1 of this calculation incorporated decay for a release taking place one hour after reactor shutdown. This was done to create continuity between the two methodologies present.2.0

SUMMARY

OF RESULTS The results of the calculations for the radiation monitors specified in the STP EAL Basis Document and are listed in Table 2.1, below.Table 2.1: Summary of Calculation Results Emergency Action Level RT-8010B, Unit Vent (pCi/sec)RT-8046 through 8049, Main Steam Lines (AtCi/cc)RU1 Unusual Event Hand Calculation 1.40E+05 5.OOE-02 STAMPEDE N/A* N/A*RA1 Alert Hand Calculation 1.57E+06 4.03E+00 STAMPEDE 2.50E+06 4.50E+00 RS1 Site Area Emergency Hand Calculation 1.57E+07 4.03E+01 STAMPEDE 2.50E+07 4.50E+O1 RG1 General Emergency Hand Calculation 1.57E+08 4.03E+02 STAMPEDE 2.50E+08 4.50E+02*STAMPEDE was not used to determine the threshold for RU1. Reference 5.10 indicates that the ODCM methodology should be used to determine the threshold value.This calculation will be used to establish the threshold values for abnormal radiation based emergencies in the STP EAL Technical Basis document. I CALC. NO. STPNOC013-CALC-002 E NERC0 N Radiological Release Thresholds REV. 2 for Emergency Action Levels Excelienre--Every ptoject Every doy f PAGE NO. 5 of47 3.0 METHOD OF ANALYSIS Previously, STAMPEDE was used to calculate the Emergency Action Level threshold values for effluent releases. A hand calculation will verify the STAMPEDE calculations. The hand calculation is described in Attachment 1 of this document STAMPEDE conforms to the requirements of STP Procedure OPGP07-ZA-0014, Software Quality Assurance Program. STAMPEDE was run at STP on an STP computer and under the supervision of an ENERCON employee with access to the STP site as a critical worker.4.0 INPUTS 4.1 Per NEI 99-01, Revision 6, Initiating condition AUI, EAL 1, the Notice of Unusual Event initiating condition is a release of gaseous or liquid radioactivity greater than two times the ODCM limit for sixty minutes or longer (Reference 5.10).4.2 The ODCM offsite dose limit is exceeded if the Xe-133 release concentration exceeds 7.41E-04[tCi/cc (Reference 5.6).4.3 The Unit Vent flow rate is 9.4E+07 cc/sec (Reference 5.1).4.4 The main steam line pressure and PORV choke flow rate are 1285 psig and 1.05E+06 lbm/hr, respectively (Reference 5.2).4.5 The specific volume of saturated steam at 1285 psig is 0.338 ft 3/lbm (Reference 5.3).4.6 The release concentration is varied to find the release concentration which correlates to each emergency action level. Emergency action levels are taken from NEI 99-01, Revision 6 (Reference 5.10) for initiating conditions AAI, ASI and AGI. EAL I is the EAL of interest in each initiating condition. The doses at the Site Boundary that correlate to the threshold concentrations are listed in Table 4.1.Table 4.1 EAL Offsite Dose Initiating Conditions Alert Site Area General TEDErem 100 mrem 1000 mrem Thyroid CDE 50 mrem 500 mrem 5000 mrem

5.0 REFERENCES

5.1 Offsite

Dose Calculation Manual, Revision 17, March 2011 5.2 Main Steam PORV Capacity Verification MC05591, Revision: 1 5.3 NIST Steam Tables, 2011 5.4 OERPO1-ZV-IN01, Emergency Classification Draft Revision 10 5.5 OERPOI-ZV-TPO1, Offsite Dose Calculations, Revision 21 5.6 STP Calculation NC-9012, CRMS Rad Monitor Setpoints, Revision 7 5.7 STP Calculation NC-901 1, Revision 2 5.8 STAMPEDE Computer Program, Revision 7.0.3.3 5.9 STAMPEDE Users Manual 5.10 NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors 5.11 OPGP07-ZA-0014 Quality Assurance Program 5.12 ITWMS Call Number 1000010987 Design Document, Revision 0 CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds CAV. N for Emergency Action Levels E ee- proe ey Y PAGE NO. 6 of 47 6.0 ASSUMPTIONS 6.1 Unit Vent Noble Gas Monitor To be consistent with the ODCM methodology, the unit vent release is assumed to be entirely Xe-133. The unit vent noble gas monitor is calibrated to Xe-133 (Reference 5.1) therefore; the monitor reading accurately reflects the Xe-133 release magnitude. To be consistent with ODCM methodology, the main steam line release is assumed to be entirely Xe-133. The noble gas monitor is calibrated to Xe-133 (Reference 5.6).6.2 Release Duration Per Reference 5.10, Sections IC AA1, ASI, and AGI developer notes, the release should be assumed to last one hour.6.3 Release following Reactor Shutdown The release initiates one hour after reactor shutdown. While a release initiating at reactor shutdown is likely, significant decay of short lived nuclides occurs during the migration time. A release at reactor shutdown would have a significantly higher activity at the monitor location than at the reception site. It is important for the threshold to not be calculated at shutdown as this would create a very high threshold which would not be appropriate for releases which occur shortly after shutdown. One hour after reactor shutdown is sufficient time to decay short lived nuclides and create a conservative threshold.

6.4 Source

Term Per Reference 5.1, any unit vent release with increased RCS activity and no core melt should be calculated using the gap inventory. Therefore, the gap inventory is used for all unit vent releases.Per Reference 5.1, for a main steam line release following a steam generator tube rupture it is appropriate to use an inventory of noble gases plus 0.2% iodine. A steam generator tube rupture is the only scenario which would create significant offsite doses through a main steam line release.6.5 Default STAMPEDE Input Values Reference 5.10 developer notes for initiating conditions AA1, AS I and AGI suggest using the ODCM or the site's emergency dose assessment methodology. STAMPEDE is used for emergency dose assessment. Per Reference 5.1, when actual meteorology is not available, the default STAMPEDE values should be used. Had the ODCM methodology been used, the 500 hour peak x/Q value would be used which is less conservative than the X/Q value produced by STAMPEDE using default meteorological conditions. Therefore, the use of STAMPEDE default values provides a more conservative estimate than that of the alternative method outlined in Reference 5.10.6.6 Average Effluent Concentration (X/Q)The same X /Q is used for the unit vent and main steam line release. Reference

5. I applies the same unit vent x/Q to Units 1 and 2 which would also be applicable to the main steam line. All releases are considered to be ground level releases.

CA LC. NO. STPNOCOI3-CA LC-002 ENER C N Radiological Release Thresholds for Emergency Action Levels REV. 2 Excee very proec. ,ev doy PAGE NO. 7 of47 7.0 STAMPEDE CALCULATIONS

7.1 Unusual

Event -RU I.7.1.1 Unit Vent Monitor AU I recommends declaring an unusual event due to a release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer (Reference 5.10).STP sets the ODCM limit at 7.41 E-04 giCi/cc (Reference 5.6, pg. 16). Two times the limit would be 1.48E-03 ltCi/cc. The threshold is listed in giCi/sec so that variations in flow rate do not change the threshold. The nonrmal flow rate from the unit vent is 9.4E+07 cc/sec (Reference 5.1).ICL\ cC \~C, Concentration ) *Flow Rate~- Release Rate ((1.48E -03) (&)*(9.4E + 07) (ec) (sec) 0 Equation 7.1.1.1 7.1.2 Main Steam Line Monitor The ODCM does not calculate a release corresponding to allowable limits for the main steam line monitors. Since the unit vent release calculated in the ODCM was assumed to be primarily Xe-133, the assumption is made in the ODCM that other noble gases and iodine may be ignored in the calculation. This assumption is equally justifiable for the main steam line and the same limiting release will be used.The magnitude of the release calculated for the unit vent Unusual Event applies to the main steam lines as well. The main steam line PORV's will create a dose exceeding two times the ODCM limit by releasing 1.4E+05 ptCi/sec of activity which is equivalent to the release from the unit vent.The steam lines hold saturated steam at 1285 psig, per Reference 5.2, which has a specific volume of 0.338ft 3/Ibm (Reference 5.3). The PORVs will release the steam at 1.05E+06 lbm/hr per Reference 5.2. This creates a set flow rate of steam from the main steam lines of 2.79E+06 cc/sec as shown below.F

• Density (lbm 28316.846 -ff + 3600 ( e sc hreC CCm)( th e 1.05E" + 06 (-=--/* 0.338 (-- *28316.846

+ 3600 2.79E +06-m)' f t3Y hr)sec Equation 7.1.2.1 CALC. NO. STPNOCOI3-CALC-002 NRadiological Release Thresholds 2 for Emergency Action Levels Excelence--Eve yproject. Ever day PAGE NO. 8 of47 Since the flow rate is set, the concentration will determine the limit. Equation 7.1.2.2 solves for the limiting concentration of 5.00E-02 /aCi/cc as shown below.Lim iting Release (I --lCI L ( Ci Limiting Concentration -Release Rate (-e-).1.40

• 105 (Ci)]1.4
• o~ ~tC) -5.OOE -02 (IICi)2.79
• 106 c (.sec)Equation 7.1.2.2 7.2 Alert, Site Area and General Emergencies

-RA1, RS 1, RG 1 7.2.1 Unit Vent Monitor Input The Alert EAL is set to 10 mrem TEDE and 50 mrem Thyroid CDE per Reference 5.10.The emergency offsite dose calculation software STAMPEDE was used to calculate the release which corresponds to this dose. A release concentration correlating to the EAL threshold value was calculated by varying the input. The following assumptions and inputs were used for the calculation as described in Sections 4.0 and 6.0.* Release begins one hour after reactor trip* Release lasts for one hour" Gap inventory source term" Default STAMPEDE input values o Windspeed = 13.2 mph o Stability class D Results Given a monitored unit vent release of 2.50E+06 pCi/sec., the Thyroid CDE is 51 mrem/hr at the closest portion of the site boundary and the EAL Initiating Condition is exceeded.Threshold values for the Site Area Emergency and General Emergencies are multiples of 10 and 100 of the Alert. Since the Correlation between release concentration and dose is linear, threshold values for the steam line monitors are 2.50E+07 and 2.50E+08 ptCi/sec for the SAE and GE respectively. Both are also limited by Thyroid CDE. Additional STAMPEDE iterations were performed to confirm this and are attached.The input and output files can be found at the end of this document in Attachment

3.

CA LC. NO. STPNOCO I3-CA LC-002 E N E R C O N Radiological Release Thresholds REV. 2 for Emergency Action Levels REV. Evefyay. t PAGE NO. 9 of 47 7.2.2 Main Steam Line Monitor Input A release concentration correlating to the EAL threshold value was calculated by varying the input. The following assumptions and inputs were used for this calculation as described in Sections 4.0 and 6.0." Release begins at reactor trip* Release lasts for one hour* Noble gas + iodine with 0.2% iodine source tern* Default STAMPEDE input values o Windspeed = 13.2 mph o Stability class D Results Given a monitored main steam line release of 4.5 [tCi/cc, the Thyroid CDE is 50 mrem/hr and the EAL Initiating Condition is exceeded.The input and output files can be found at the end of this document in Attachment

3.7.3 Threshold

values for the Site Area Emergency and General Emergencies are multiples of 10 and 100 of the Alert. Since the correlation between release concentration and dose is linear, threshold values for the steam line monitors are 45 and 450 itCi/cc for the SAE and GE respectively. Both are also limited by Thyroid CDE. Additional STAMPEDE iterations were performed to confirm this and are attached. Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 ENERCO N for Emergency Action Levels REV. 2 Excel!ence--Everyproject Every doy Attachment 1 PAGE NO. 10 of47 Attachment 1 -Hand Calculations 1.0 OBJECTIVE/SCOPE Each release calculated using STAMPEDE in the main document is calculated by hand in this attachment and the results compared to STAMPEDE.2.0

SUMMARY

OF RESULTS Table 2.1 is displayed again below showing the results from all the calculations. The minor difference is due to STAMPEDE using decay factors over a one hour period after shutdown. This also accounts for the change in the limiting dose being TEDE in the hand calculations and Thyroid CDE in the STAMPEDE calculations. The accuracy of the hand calculation is considered sufficient and recommended for use in Emergency Action Levels.Table 2.1 Results Emergency Action Level RT-8010b, Unit Vent (jiCi/sec) RT-8046 through 8049, Main Steam Line (RCi/cc)RU1 Unusual Event Hand Calculation 1.40E+05 5.OOE-02 STAMPEDE N/A N/A RAI Alert Hand Calculation 1.57E+06 4.03E+00 STAMPEDE 2.50E+06 4.50E+00 RS1 Site Area Emergency Hand Calculation 1.57E+07 4.03E+O1 STAMPEDE 2.50E+07 4.50E+O1 RG1 General Emergency Hand Calculation 1.57E+08 4.03E+02 STAMPEDE 2.50E+08 4.50E+02 3.0 METHOD OF ANALYSIS Using the limiting dose at the site boundary, the release is back calculated using atmospheric dispersion models. The X/Q value used is calculated from Regulatory Guide 1.145, Atmospheric Dispersion Modelsfor Potential Accident Consequence Assessments at Nuclear Power Plants. Rather than using the most conservative meteorology, average meteorological conditions are used as inputs Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 EN ER CO N for Emergency Action Levels REV. 2 verydy. Attachment 1 PAGE NO. II of 47 to most closely agree with STP emergency dose assessment methodology per the ODCM and STAMPEDE. Assumed nuclide inventories are taken from Reference 5.15. The dose conversion factors are taken from Reference 5.2. A release concentration is used to find an initial projected dose at the Site Boundary. Using the projected dose at the Site Boundary, the release concentration is scaled to find the limiting dose for each EAL.4.0 INPUTS* The Unit Vent flow rate is taken from the Offsite Dose Calculation Manual; Revision 17, March 2011 and is 9.44E+07 cc/sec." The main steam line pressure and PORV choke flow rate were taken from Reference 5.5 and are 1285 psig and 1.05E+06 lbm/hr respectively.

• The specific volume of saturated steam at this pressure is taken from the NIST steam tables and is 0.338 ft 3/lbm.* The release concentration is varied to find the release concentration which correlates to each emergency action level dose. Emergency action level doses are taken from NEI 99-01 Revision 6 for initiating conditions AAI, ASI and AG1. EAL I is the EAL of interest in each initiating condition.

The limiting doses are listed in Table 4.1. NEI 99-01 Revision 6 states that these values are based on fractions of the Environmental Protection Agencies Protective Action Guidelines (EPA PAGs) and the General Emergency represents the protective action values recommended by the EPA.Table 4.1 EAL Thresholds Alert Site Area General TEDE 10 mrem 100 mrem 1000 mrem Thyroid CDE 50 mrem 500 mrem 5000 mrem" A release lasting one hour is selected per NEI 99-01 Revision 6 developer notes.* Atmospheric dispersion factors are calculated per Regulatory Guide 1.145 (Reference 5.1). The reactor building dimensions used as inputs for this calculation are taken from Reference 5.13.* Nuclide mixes are taken from Reference 5.15, which is the source document for the nuclide mixes used in STAMPEDE. The release mixes are a gap release and noble gases plus 0.2% iodine which are listed below. Each nuclide mix was normalized to one so it could be scaled to various release activities. Radiological Release Thresholds CALC. NO. STPNOCO 1 3-CALC-002 U EN ER CO N for Emergency Action Levels REV. 2 Excellence--eyeproject. Everydoay Attachment I PAGE NO. 12 of47 Table 4.2 Gap Inventory Nuclide 1-131 1-132 1-133 1-134 1-135 Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-131m Xe-133m Xe-133 Xe-135m Mix 1.I1OE+05 1.50E+05 2.20E+05 2.40E+05 2.OOE+05 1.30E+06 2.90E+06 3.70E+05 5.50E+06 7.80E+06 9.50E+06 1.1OE+05 6.80E+05 2.20E+07 4.20E+06 Normalized

1. 1 2E-03 1.53E-03 2.25E-03 2.45E-03 2.05E-03 1.33E-02 2.97E-02 3.78E-03 5.62E-02 7.98E-02 9.72E-02 1.1 2E-03 6.95E-03 2.25E-0 1 4.30E-02 I Nuclide Xe-135 Xe-137 Xe-138 Cs-134 Cs-137 Te132 Mo99 RulO3 Ru106 Zr95 Lai40 Ce144 Ce-141 Sr89 Sr9o Mix Normalized 5.50E+06 5.62E-02 1.90E+07 1.94E-0I 1.80E+07 1.84E-01 3.70E+/-01 3.78E-07 2.90E+01 2.97E-07 4.80E+00 4.91E-08 1.22E+O1 1.25E-07 8.80E-03 9.OOE-1 1 2.90E-03 2.97E- I1 1.1OE-02 1.12E-10 1.90E-02 1.94E-10 7.40E-03 7.57E-1 I 1.OOE-02 1.02E-10 6.40E-02 6.55E-10 3.20E-03 3.27E-1 1 Table 4.3 Noble Gases+0.2%

Iodine Inventory Nuclide 1-131 1-132 1-133 1-134 1-135 Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Xe-137 Xe-138 Kr-83m Kr-85 Kr-85m Kr-87 Kr-88 Kr-89 Mix 6.1OE-02 8.61 E-02 1.OOE-01 1.86E-02 2.73 E-0 I 2.80E+00 2.40E+02 4.20E+00 7.60E+00 4.OOE-0 1 1.60E-01 5.80E-0 I 3.70E-01 7.60E+00 1.50E+00 9.80E-01 2.80E+00 8.40E-02 Normalized 2.26E-04 3.19E-04 3.72E-04 6.92E-05 1.0 1E-03 1.04E-02 8.90E-01 1.56E-02 2.82E-02 1.48E-03 5.93E-04 2.15E-03 1.37E-03 2.82E-02 5.56E-03 3.63E-03 1.04E-02 3.12E-04 (Reference 5.2) are listed in Tables 4.4 The dose conversion factors taken from EPA 400R92001 and 4.5 below. Radiological Release Thresholds CALC. NO. STPNOCOI3-CALC-002 SE ER C O N for Emergency Action Levels REV. 2 Ex'ellence--Veryproject Fverydoy. Attachment I PAGE NO. 13 of47 Table 4.4 TEDE Dose Conversion Factors Nuclide 1-131 1-132 1-133 1-134 1-135 Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-131m Xe-133m Xe-133 Xe-135m Dose Conversion Factor (rem per uCi*hr/cc) 5.30E+04 4.90E+03 1.50E+04 3.1OE+03 8.1OE+03 9.30E+01 1.30E+00 5.1OE+02 1.30E+03 1.20E+03 4.9 1.70E+O1 2.OOE+O 1 2.50E+02 Nuclide Xe-135 Xe-137 Xe-138 Cs-134 Cs-137 Tel32 Mo99 Ru103 RulO6 Zr95 Lal40 Ce144 Ce-141 Sr89 Sr90 Dose Conversion Factor (rem per uCi*hr/cc) 1.40E+02 1.1OE+02 7.20E+02 6.30E+04 4.1OE+04 1.20E+04 5.20E+03 1.30E+04 5.70E+05 3.20E+04 1.1OE+04 4.50E+05 1.1OE+04 5.OOE+04 1.60E+06 Table 4.5 Thyroid CDE Dose Conversion Factors Nuclide 1-132 1-133 1-134 1-135 Thyroid CDE DCF (rem per uCi*hr/cc) 1.30E+06 7.70E+03 2.20E+05 1.30E+03 3.80E+04 The unit vent noble gas monitor energy efficiency by nuclide is taken from Offsite Dose Calculation Manual (Reference 5.3). The values are relative to Xe-133 efficiency since the monitor is calibrated to Xe-133. Table 4.6 displays the energy efficiency by nuclide relative to Xe-133. Radiological Release Thresholds CALC. NO. STPNOCOI3-CALC-002 E N ER CO N for Emergency Action Levels REV. 2 ExceP, p .cvFeyday Attachment I PAGE NO. 14 of47 Table 4.6 Energy Efficiency Relative to Xe-133 Efficiency Relative to Xe-133 Nuclide (U/Ceuian....................... ....... ..... ............ .(" C ~ s e ~ " e Kr-83m *Kr-85m 1.9 Kr-85 2.4 Kr-87 2.8 Kr-88 2.3 Kr-89 2.8 Xe-131m 0.015 Xe-133m 0.14 Xe-133 I Xe-135m 0.042 Xe-135 2.5 Xe-137 2.8 Xe-138 2.8*There is no relative efficiency available for Kr-83m.omission.t --Assumption

6.4 further

justifies the Table 4.7 Nuclide Half Lives Nuclide Half Life (hr)1-131 1.93E+02 1-132 2.38E+00 1-133 2.03E+01 1-134 8.77E-01 1-135 6.61E+00 Kr-83m 1.83E+00 Kr-85m 4.48E+00 Kr-85 9.40E+04 Kr-87 1.27E+00 Kr-88 2.84E+00 Kr-89 5.1OE-02 Xe-131m 2.83E+02 Xe-133m 5.42E+01 Xe-133 1.27E+02 Xe-135m 2.60E-01 Nuclide Half Life (hr)".Xe-135 Xe-137 Xe-138 Cs-134 Cs-137 Te132 Mo99 Ru103 Rul06 Zr95 Lal40 Ce144 Ce-141 Sr89 Sr90 9.08E+00 6.38E-02 2.36E-01 1.80E+04 2.60E+05 7.79E+0 I 6.62E+0 I 9.44E+02 8.84E+03 1.55E+03 4.03E+01 6.82E+03 7.77E+02 1.21 E+03 2.50E+05 0 The half-lives are taken from Reference 5.15 which lists the input data used by STAMPEDE. Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 E N E R C O N for Emergency Action Levels RE\'.2 Eaceferxp-Every ptojea. Every day Attachment I PAGE NO. 15 of 47

5.0 REFERENCES

5.1 Regulatory

Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Revision 1, November 1982.5.2 EPA 400R92001, Manual of Protective Action Guides and Protective actions for Nuclear Incidents, Revision 1, May 1992: 5.3 Offsite Dose Calculation Manual, Revision 17, March 2011.5.4 TGX/THX 3-1, Revision 5., Westinghouse Radiation Analysis Manual.5.5 MC05591, Main Steam PORV Capacity Verification, Revision 1.5.6 NIST Steam Tables, 2011.5.7 OERPO I-ZV-INO1. Emergency Classification, Revision 10.5.8 0ERP01-ZV-TPO1, Offsite Dose Calculations, Revision 21.5.9 STP Calculation NC-9012, Process and Effluent Radiation Monitor Set Points, Revision 7 5.10 STP Calculation NC-901 1, CRMS Rad Monitor Setpoints, Revision 2.5.11 STAMPEDE Computer Program, Revision 7.0.3.3.5.12 STAMPEDE Users Manual 5.13 STP Drawing 6C189N5007, General Arrangement Reactor Containment Building, Revision 6 5.14 NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors 5.15 ITWMS Call Number 1000010987 Design Document, Revision 0 6.0 ASSUMPTIONS

6.1 Release

lasts for one hour Per NEI 99-01 (Reference 5.14), IC AA1, AS], AGI developer notes,.the release should be assumed to last one hour.For this to be true for the main steam line, it is assumed that the PORV is open for one hour. To calculate the most limiting case, it is assumed that the maximum flow possible is being released friom the PORV.6.2 Nuclide mix Per 0ERP0I-ZV-TPO1, Offsite Dose Calculations (Reference 5.8) any unit vent release with increased RCS activity and no core melt should be calculated using a gap inventory. It is conservative to assume an increased RCS activity and not within the intended scope of the relevant initiating conditions to assume core melt. Therefore, a gap inventory is used for all unit vent releases.Per 0ERP0I-ZV-TPO1, Offsite Dose Calculations (Reference 5.8) for a main steam line release following a steam generator tube rupture it is appropriate to use an inventory of 100 percent noble gases plus 0.2 percent iodine. STAMPEDE's source mix for noble gases plus iodine is defined by the noble gas levels from the coolant mix plus the iodine levels from the coolant mix scaled to a percentage of total noble gas (Reference 5.15). A definition of 0.2 percent iodine means the summation of iodine levels is equivalent to 0.2 percent of the noble gas levels. Since a steam generator tube rupture releasing through the PORVs is the only steam generator tube rupture Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 F4 E N E R C O N for Emergency Action Levels REV. 2 Exceflence--Evetyptont. E'ety day Attachment 1 PAGE NO. 16 of47 scenario which would create offsite doses large enough to meet or exceed the EALs, this assumption is made.6.3 Atmospheric Dispersion NEI 99-01 (Reference 5.14) developer notes for initiating conditions AA1, ASI and AGI suggest using the ODCM or the site's emergency dose assessment methodology. Per OERPO 1 -ZV-TPO 1, Offsite Dose Calculations (Reference 5.8), when actual meteorology is not available, the default STAMPEDE values should be used. The default STAMPEDE values assume a stability class D for atmospheric dispersion and a windspeed of 13.2 mph. These values were used as inputs for the atmospheric dispersion calculation. It is clear that STAMPEDE uses the same method for calculating atmospheric dispersion factor (X/Q) outlined in section 7.1.1 of this Attachment. However, STAMPEDE does not follow the same logic in selecting the appropriate result from the three calculations. The STAMIPEDE value printed in the results found in attachment 3 is consistent with the largest of the three hand calculated X/Q values. This suggests that STAMPEDE simply selects the largest of the three X/Q values resulting in a much more conservative estimate. This calculation will deviate from the recommendations of Regulatory Guide 1.145 and conform to the methodology STAMPEDE uses.The close proximity of all release points allows for a single atmospheric dispersion coefficient to be used. This assumption is also made by STAMPEDE.6.4 Exposure Pathways The dose conversion factors used in Tables 4.4 and 4.5 represent a summation of dose conversion factors for external plume exposure, inhalation from the plume and external exposure from deposition. Because the dose estimations are used for implementing early phase protective actions, conversion factors using limited pathways are appropriate. The EPA does not provide a dose conversion. factor for Kr-83m. Because the PAGs are based on EPA dose calculations, it is appropriate to only use the nuclides for which dose conversion factors are provided. Additionally, Kr-83mn represents only 1.33% of the nuclide inventory activity and its exclusion would not significantly affect the final dose.6.5 Decay The release initiates one hour after reactor shutdown. While a release initiating at reactor shutdown is likely, significant decay of short lived nuclides occurs during the migration time. A release at reactor shutdown would have a significantly higher activity at the monitor location than at the reception site. It is important for the threshold to not be calculated at shutdown as this would create a very high threshold which would not be appropriate for releases which occur shortly after shutdown. One hour after reactor shutdown is sufficient time to decay short lived nuclides and create a conservative threshold. Decay is incorporated for one hour friom reactor shutdown as well as migration time. Half-lives are taken from Reference 5.15. Migration time is assumed to be the reciprocal of the wind speed. Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 E N E R C O N for Emergency Action Levels REV. 2 Exceieernce--veryproject. Fveryday Attachment 1 PAGE NO. 17 of 47 7.0 HAND CALCULATIONS 7.1 Unit Vent Monitor 7.1.1 X/Q The atmospheric dispersion factor, X/Q, determines the change in concentration between the unit vent discharge and the dose reception site. This value is based on meteorological conditions and will vary with wind speed and stability class. The ODCM uses the highest annual average X/Q value at the site boundary which is 5.3 E-06 sec/r 3.However, for an accident related release STAMPEDE is used rather than the ODCM. STAMPEDE uses real time, user entered., or default meteorological conditions to calculate the X/Q for a specific accident. Default values will be used as inputs into the Regulatory Guide 1.145 method for calculating X/Q as described below. Default values are identified in section 6.0, Atmospheric Dispersion. For a neutral atmospheric stability class, which is the default in STAMPEDE. X/Q values can be determined through the following set of equations. X 1 Equation 7.1.1.1 X 1 Q Ulb (3cruycrz) Equation 7.1.1.2 X 1 Q U 1 0 o Yuz Equation 7.1.1.3 Where X/Q = relative concentration (sec/m^3)7C =3.14159 U 1 0 = windspeed at 10 meters above plant grade (m/s)ury = lateral plume spread (in), a function of atmospheric stability and distance, determined from Regulatory Guide 1.145 Figure 1 UZ = vertical plume spread (m), a function of atmospheric stability and distance, determined from Regulatory Guide 1.145 Figure 2 Vy = (M -1)aysoom + ay = lateral plume spread wvith meander and building wake effects (m), a function of atmospheric stability, windspeed U 1 0 , and distance; M is determined from Regulatory Guide 1.145 Figure 3 A = the smallest vertical-plane cross-sectional area of the reactor building (mA2).taken firom Reference 5.13 and shown below Radiological Release Thresholds CALC. NO. STPNOCOI3-CALC-002 NE CE 0 for Emergency Action Levels REV. 2 Eveirerre-Eeyproyrrt ery day Attachment I PAGE NO. 18 of 47 Figure 7.1.1.1: Reactor Building Dimensions EL 241"-0i Assuming the reactor building cross section to be a perfect rectangle and half sphere, the variables are defined as follows;U 1 0 = 13.2 mph = 5.9 m/s uy = 1200 in a, = 4.2 m Ey = (M -1)uy8o0m + -y ; M=I --y = 1200 l m A = (135'* 158')+ ( 2)= 31128.37 The three equations become;x Q 1 5.9 (7r1200

• 4.2 + 31128.37) 5.398
• 106 Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 , E N E R C O N for Emergency Action Levels REV. 2 Excelerce-Evety project. Every doy. Attachment 1 PAGE NO. 19 of 47= 1= 3.568*10-6 Q 5.9(37r
• 1200
• 4.2)X 1 x 1 -1.07
• 10-5 Q 5.9
• 7r * [(1 -1)ay800oo

+ 1200]

• 4.2 To select the appropriate X/Q value, the first two X/Q values should be compared and the higher value selected.

This value is then compared with the third X/Q value and the lower of those two is the appropriate X/Q value. The appropriate X/Q is 5.3 9E-06 sec/m 3 for default meteorological conditions by the methodology recommended in Regulatory Guide 1.145.This calculated value is very similar to the ODCM highest average value of 5.3E-06 sec/m 3 which was not selected for use. Additionally, the value shown in the STAMPEDE output file at one mile is 1.032E-05 sec/m 3.This suggests that STAMPEDE uses the same methodology and simply selects the largest atmospheric dispersion value to remain conservative. This methodology will be replicated and 1.07E-05 will be used as the X/Q.7.1.2 Nuclide Inventory As previously stated, a gap mix is appropriate for this problem. The gap mix is taken from Reference 5.15 which is used as the source term for STAMPEDE nuclide mixes.The mix was then normalized so they could be scaled to the varying emergency classifications. The values for the normalized inventory can be found in Table 4.2.7.1.3 Dose Conversion Factors As stated in NEI 99-01 (Reference 5.14) developer notes, the purpose of dose projections is to check if the Environmental Protection Agencies Protective Action Guidelines (EPA PAGs) have been exceeded. The dose conversion factors provided by the EPA in EPA 400R92001 are used. These dose conversion factors account for external plume exposure, inhalation from the plume, and external exposure from deposition and are listed in Tables 4.4 and 4.5, and taken from Tables 5-1 and 5-2 in EPA 400R92001 (Reference 5.2).The EPA does not provide a dose conversion factor for Kr-83m. This nuclide contributes 1.33% of the inventory activity. The lack of this nuclide's contribution to the final dose will not significantly affect the outcome.7.1.4 Decay Time One hour of decay is incorporated to the monitor response due to the release initiating one hour after reactor shutdown. Decay is also incorporated for the duration of the migration time. The total decay time is one hour plus the reciprocal of wind speed, or 1.07575 hours.7.1.5 Dose Calculations Radiological Release Thresholds CALC. NO. STPNOCOI3-CALC-002 E N E R C O for Emergency Action Levels REV. E Exce.ence--Evev, project. Fvery day. Attachment 1 PAGE NO. 20 of 47 The dose rate at the site boundary is calculated using Equation 7.1.5.1.Sn 1.07575/ -F Ci*0.5 T 1/2i *DCF 1 Equation 7.1.5.1 Where D = dose rate per hour at the site boundary-atmospheric dispersion coefficient as calculated in section 7.1.1 Q F = unit vent flow rate Ci= concentration of nuclide i at the time of shutdown 1.07575 = the total decay time of interest from section 7.1.4 TI/2i = the half-life of nuclide i DCF 1 = the dose conversion factor for nuclide i listed in tables 4.4 and 4.5 The total concentration of the nuiclides is varied to find the dose rate of interest.Beginning with an arbitrary release concentration of I [tCi/cc, the dose rate is calculated. Since the dose is linearly correlated to concentration, the release concentration may be scaled to find the dose rate of interest.The Alert EAL is 10 mrem TEDE or 50 mrem Thyroid CDE. Using the above method to calculate TEDE with the appropriate conversion factors, a limiting release rate of 2.33E+06 gCi/sec from the unit vent results in 5.7 mrem TEDE. Using the calculated release rate to find Thyroid CDE with the appropriate conversion factors, the same release results in a 50 mrem Thyroid CDE at the site boundary. Thus, 2.33E+06 pCi/sec is the limiting release rate based on the 50 mrem Thyroid CDE EAL initiating condition. The limiting release rate threshold values for the Site Area Emergency and General Emergencies are multiples of 10 and 100 of the Alert release rate threshold value.These calculations can be found in Attachment 2.7.1.6 Monitor Response The unit vent noble gas monitor is calibrated to Xe-133. Monitor efficiencies relative to Xe-133 by nuclide are listed in ODCM Table B3-2. To find the monitor reading associated with each limiting release, the noble gas concentrations must be multiplied by the monitor response and summed. Table 4.6 shows the indicated response of the unit vent noble gas monitor by nuclide and Equation 7.1.5.2 shows how the monitor response was calculated. Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 E N E R C O N for Emergency Action Levels REV. 2 Excel ence-Every project. very doy Attachment I PAGE NO. 21 of47 Monitor Response =

• Rej Equation 7.1.5.2 Where C 1 = concentration of nuclide i (pCi/cc)Re 1 = monitor response to nuclide i (.tCi/cc)x_133 equivalent In the case of an Alert, the 2.33E+06 ltCi/sec release rate will read as 1.57E+06 pCi/sec on the monitor. Kr-83m does not have an indicated monitor response coefficient.

Because Kr-83m is only 1.33% of the noble gases and does not contribute to the dose calculation, its exclusion is acceptable. This again is a linear correlation and the SAE and GE scale by factors of 10 and 100 respectively. These calculations can be found in Attachment 2.7.2 Main Steam Line Monitors 7.2.1 X/Q Since the atmospheric dispersion is independent of nuclide inventory or release rate and the close proximity of thereleases, the X/Q value will be the same for a main steam line release as it is for a unit vent release. This assumption is also taken by STAMPEDE and outline in Assumption 6.3.7.2.2 Nuclide Inventory Per OERPO1-ZV-TPO0, if the release path is the main steam line with a steam generator tube rupture, the nuclide inventory should be 100% noble gas from the reactor coolant and the reactor coolant iodine mix scaled to equal 0.2 percent of the total noble gas.The secondary steam concentration for noble gases and iodine after a steam generator tube rupture are taken from Reference 5.15. Values for the reactor coolant inventory are listed in Table 4.3. All of the noble gases are used and the iodine concentration from the coolant inventory is scaled to total 0.2% of the total noble gas mix. These inventories are then normalized to one. These values are listed in Table 4.3.7.2.3 Dose Conversion Factors The dose conversion factors used are found in Tables 4.4 and 4.5, taken from Tables 5-1 and 5-2 in EPA 400R92001. Radiological Release Thresholds CALC. NO. STPNOCO13-CALC-002 E N E R CO N for Emergency Action Levels REV. 2 ExceIence-Evetypfojev. Evetydoy Attachment I PAGE NO. 22 of47 7.2.4 Decay Time One hour of decay is incorporated to the monitor response due to the release initiating one hour after reactor shutdown. Decay is also incorporated for the duration of the migration time. The total decay time is one hour plus the reciprocal of wind speed, or 1.07575 hours.7.2.5 Dose Calculations Equation 7.1.5.1 applies to the release from the main steam lines. The main steam line flow rate is used instead of the unit vent flow rate for the value F. The main steam line flow rate was calculated in Equation 7.1.2.2 of the STAMPEDE CALCULATIONS section of this document as 2.79E+06 cc/sec.The Alert EAL threshold is 10 mrem TEDE or 50 mrem Thyroid CDE at the site boundary (Table 4.2). Using the method in Equation 7.1.5.1 to calculate TEDE with the appropriate conversion factors, a concentration at time of shutdown of 4.10 i.tCi/cc would result in 6.89 mrem TEDE at the site boundary if the steam line PORV was open for an hour. Using the same steam line concentration to calculate Thyroid CDE results in 50 mrem Thyroid CDE at the, site boundary.The steam line concentrations at the time of shutdown for the Site Area Emergency and General Emergencies are multiples of 10 and 100 of the Alert. Since the correlation between release concentration and dose is linear, values for the steam line concentration at time of shutdown are 41.0 and 410 jiCi/cc for the SAE and GE respectively. Both are also limited by Thyroid CDE.These calculations can be found in Attachment 2.7.2.6 Monitor Response Because the main steam line monitor is adjacent to the main steam line, significant shielding takes place between the source and monitor. STP calculation NC-90 11 Revision 2 calculates a conversion factor for the main steam lines for a noble gas inventory which is incorporated into the monitor readout. No monitor response needs to be calculated. The concentration of the main steam line one hour after shutdown given a concentration of 4.10 ltCi/cc at the time of shutdown is 4.03 ptCi/cc. This calculation is also found in Attachment

2. Additionally, the monitor readings for the SAE and GE one hour after shutdown are 40.3 and 403 piCi/cc respectively.

These values are the thresholds for the main steam line monitor. Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 0 E NERCON for Emergency Action Levels REV.2 Excelnce-EEvery pq Fery ody Attachment 2 PAGE NO. 23 of 47 Table A2-1: Unusual Event Emergency Calculations 1.40E+05 I 2.79E+06 I 5.OOE-02 Table A2-2: Input Values for Calculations Radiological Release Thresholds CALC. NO. STPNOCOI3-CALC-002 I EN ERCO N for Emergency Action Levels REV.2 Excelence-Evenyproject Every day Attachment 2 PAGE NO. 24 of 47 Table A2-3: Calculations for Boundary Concentrations and TEDE dose due to Unit Vent Release 1-131 1-132 1-133 1-134 1-135 Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 1.1OE+05 1.50E+05 2.20E+05 2.40E+05 2.OOE+05 1.30E+06 2.90E+06 3.70E+05 5.50E+06 7.80E+06 9.50E+06 1.1OE+05 6.80E+05 2.20E+07 4.20E+06 5.50E+06 1.90E+07 1.80E+07 1.12E-03 1.53E-03 2.25E-03 2.45E-03 2.05E-03 1.33E-02 2.97E-02..78E-03 5.62E-02 7.98E-02 9.72E-02 1.12E-03 6.95E-03 2.25 E-01 4.30E-02 5.62E-02 1.94E-0 I 1.84E-01 2.76E-05 3.77E-05 5.55E-05 6.04E-05 5.06E-05 3..28E-04 7.33E-04 9.33)E-05 1.39E-03 1.97E-03 2.40E-03 2.76E-05 1.71 E-04 5.55E-03 1.06E-03 1.39E-03 4.79E-03 4.54E-03 1.0 1E-03 1.0 1E-03 1.0 1E-03 1.0IE-03 1.0 1E-03 1.0 1E-03 1.0 1E-03 1.0 1E-03 1.0 1E-03 1.01E-03 1.0 1E-03 1.0 1E-03 1.0 1E-03 1.0 1E-03 1.0 1E-03 1.0 1E-03 1.0 1E-03 1.01E-03 2.79E-08 1.93E+02 2.78E-08 5.30E+04 3.8 1 E-08 2.38E+00 2.79E-08 4.90E+03 5.61 E-08 2.03 E+0 I 5.40E-08 1.50E+04 6.11 E-08 8.77E-0 1 2.61 E-08 3.1 OE+03 5 11 E-08 6.61 E+00 4.56E-08 8.1 OE+03 3.3 1E-07 7.40E-07 9.42E-08 1.40E-06 1.99E-06 2.42E-06 2.79E-08 1.73E-07 5.61 E-06 1.07E-06 1.40E-06 4.83E-06 4.59E-06 1.83E+00 4.48E+00 9.40E+04 1.27E+00 2.84E+00 5.1OE-02 2.83E+02 5.42E+01 1.27E+02 2.60E-0 I 9.08E+00 6.38E-02 2.36E-0 I 2.21 E-07 6.27E-07 9.42E-08 7.79E-07 1.53E-06 1.08E-12 2.78E-08 1.71 E-07 5.57E-06 6.09E-08 1.29E-06 4.06E- I1 1.95E-07 9.30E+0 I 1.30E+00 5.10E+02 1.30E+03 1.20E+03 2.50E+02 1.40E+02 1.1OE+02 7.20E+02 5.30E+04 4.90E+03 1.50E+04 1.47E-03 1.37E-04 8.11 E-04 8.09E-05 3.70E-04 0.OOE+00 5.83E-05 1.22E-07 3.97E-04 1.99E-03 1.30E-09 1.36E-07 2.90E-06 1.11 E-04 1.52E-05 1.81E-04 4.47E-09 1.40E-04 Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 f ENERCoN for Emergency Action Levels REV. 2EC Nf_47 Excellence-Everyproject Every day, Attachment 2 PAGE NO. 25 of 47 Cs-134 Cs-137 Tel32 Mo99 Ru103 Ru 106 Zr95 Lal40 Ce 144 Ce- 141 Sr89 Sr90 3.70E+Ol 3.78E-07 9.33E-09 1.01E-03 9.42E-12 1.80E+04 9.42E-12 6.30E+04 5.93E-07 2.90E+O1 2.97E-07 7.33E-09 1.01E-03 7.40E-12 2.60E+05 7.40E-12 4.1OE+04 3.03E-07 4.80E+00 4.91E-08 1.21E-09 1.01E-03 1.22E-12 7.79E+01 1.21E-12 1.20E+04 1.45E-08 1.22E+01 1.25E-07 3.08E-09 1.01E-03 3.11E-12 6.62E+01 3.08E-12 5.20E+03 1.60E-08 8.80E-03 9.OOE-11 2.22E-12 1.01E-03 2.24E-15 9.44E+02 2.24E-15 1.30E+04 2.91E-I 1 2.90E-03 1.1OE-02 1.90E-02 7.40E-03 1.00E-02 6.40E-02 3.20E-03 2.97E- II 1.12E-10 1.94E-10 7.57E- II 1.02E-10 6.55E- 10 3.27E- I1 7.33E-13 2.76E- 12 4.79E-12 1.87E-12 2.52E-12 1.62E-1 1 8.07E- 13 1.0 1E-03 1.0 1E-03 1.0 1E-03 1.0 1E-03 1.0 1E-03 1.0 1E-03 1.0 1E-03 7.40E-16 2.79E- 15 4.83E-15 1.89E-15 2.54E-15 1.63E-14 8.15E-16 8.84E+03 1.55E+03 4.03E+01 6.82E+03 7.77E+02 1.21 E+03 2.50E+05 7.40E-16 2.79E- 15 4.75E-15 1.89E-15 2.54E-15 1.63E-14 8.15E-16 5.70E+05 3.20E+04 1.t0E+04 4.50E+05 1.10E+04 5.OOE+04 1.60E+06 4.22E-10 8.93E-1 I 5.22E- II 8.49E-10 2.79E- II 8.16E- 10 1.30E-09 Total TEDE Dose 5.77E-03 Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 10 E -E RC ON for Emergency Action Levels REV. 2ECOf_47 Excellence-E ery proec Ever ydoy Attachment 2 PAGE NO. 26 of 47 Table A2-4: Thyroid Dose Calculation for Unit Vent Release 1-131 2.78 E-08 1.30E+06 3.61 E-02 1-132 2.79E-08 7.70E+03 2.15E-04 1- 133 5.40E-08 2.20E+05 1.19E-02 1-134 2.61 E-08 1.30E+03 3.39E-05 1-135 4.56E-08 3 .80E+04 1.73E-03 Table A2-5: Unit Vent Monitor Response to Nuclide Inventory Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-131m Xe-133m Xe-133 Xe- 135m Xe-135 Xe-137 Xe-138 3.28E-04 7.33E-04 9.33E-05 1.39E-03 1.97E-03 2.40E-03 2.76E-05 1.71 E-04 5.55E-03 1.06E-03 1.39E-03 4.79E-03 4.54E-03 1.83 E+00 4.48E+00 9.40E+04 1.27E+00 2.84E+00 5.1OE-02 2.83E+02 5.42E+0 I 1.27E+02 2.60E-0 I 9.08E+00 6.38E-02 2.36E-01 2.25E 6.28E 9.33E 8.03E 1.54E 3.00E 2.76E 1.69E 5.52E 7.38E 1.28E 9.15E 2.41E-04-04 1.9-05 2.4-04 2.8-03 2.3-09 2.8-05 0.015-04 0.14-03 1-05 0.042-03 2.5-08 2.8-04 2.8 Monitor Reading: 0.00E+00 1.19E-03 2.24E-04 2.25E-03 3.55E-03 8.40E-09 4.13E-07 2.37E-05 5.52E-03 3.1OE-06 3.21 E-03 2.56E-07 6.74E-04 (uCi/cc) (uCi/sec) Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 10 N ERCO N for Emergency Action Levels REV. 2 Excellmen-Everyproject, EverydW Attachment 2 PAGE NO. 27 of 47 Table A2-6: Input for Main Steam Line Release Calculation Table A2-7: Calculations for Boundary Concentrations 1-131 1-132 1-133 1-134 1-135 Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Xe-137 Xe-138 Kr-83m Kr-85 Kr-85m Kr-87 Kr-88 Kr-89 6.I01E-02 8.6 IE-02 1.00E-0 I 1.86E-02 2.73 E-0 I 2.80E+00 2.40E+02 4.20E+00 7.60E+00 4.OOE-0 I 1.60E-01 5.80E-01 3.70E-0 1 7.60E+00 1.50E+00 9.80E-01 2.80E+00 8.40E-02 2.26E-04 3.19E-04 3.72E-04 6.92E-05 1.0 1E-03 1.04E-02 8.90E-01 1.56E-02 2.82E-02 1.48E-03 5.93E-04 2.15E-03 1.37E-03 2.82E-02 5.56E-03 3.63E-03 1.04E-02 3.12E-04 9.27E-04 1.3 1E-03 1.53E-03 2.84E-04 4.14E-03 4.26E-02 3.65E+00 6.40E-02 1.16E-01 6.07E-03 2.43E-03 8.82E-03 5.62E-03 1.16E-0I 2.28E-02 1.49E-02 4.26E-02 1 .28E-03 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.77E-08 3.90E-08 4.55E-08 8.47E-09 1.24E-07 1.27E-06 1.09E-04 1.9 1E-06 3.45E-06 1.81E-07 7.26E-08 2.63 E-07 1.68E-07 3.45 E-06 6.81 E-07 4.44E-07 1.27E-06 3.82E-08 1.93E+02 2.38E+00 2.03E+O1 8.77E-01 6.61E+00 2.83E+02 1.27E+02 5.42E+01 9.08E+00 2.60E-01 6.38E-02 2.36E-0 I 1.83E+00 9.40E+04 4.48E+00 1.27E+00 2.84E+00 5.1 OE-02 2.76E-08 2.85E-08 4.39E-08 3.62E-09 1.1OE-07 1.27E-06 1.08E-04 1.88E-06 3.18E-06 1.03E-08 6.10E-13 1.12E-08 1.12E-07 3.45E-06 5.76E-07 2.47E-07 9.79E-07 1.71E-14 5.30E+04 4.90E+03 1.50E+04 3.1OE+03 8.1 OE+03 4.9 2.OOE+01 1.70E+01 1.40E+02 2.50E+02 1.40E+02 7.20E+02 1.30E+00 9.30E+01 5.1OE+02 L .30E+03 1.20E+03 1.46E-03 1.40E-04 6.58E-04 1.12E-05 8.95E-04 6.22E-06 2.17E-03 3.20E-05 4.45 E-04 2.57E-06 8.53E- II 8.04E-06 0.OOE+00 4.49E-06 5.36E-05 1.26E-04 1.27E-03 2.05E-1 1 Total Dose 7.28E-03*Release Constant = X/Q

• duration
• release rate Radiological Release Thresholds CALC. NO. STPNOCOI3-CALC-002 EN E RCO N for Emergency Action Levels REV. 2 Exctenl --Eer, project Eecry day Attachment 2 PAGE NO. 28 of 47 Table A2-8: Main Steam Line Release Thyroid Dose Calculation 1-131 2.76E-08 1.30E+06 3.58E-02 1-132 2.85E-08 7.70E+03 2.20E-04 1-133 4.39E-08 2.20E+05 9.66E-03 1-134 3.62E-09 1.30E+03 4.71 E-06 1-135 1.1OE-07 3.80E+04 4.20E-03 4.9E0 CALC. NO. STPNOCOI 3-CALC-002 Radiological Release Thresholds FE C O N for Emergency Action Levels REV. 2 Excellence-Everyprojct, Every doy Attachment 3 PAGE NO. 29 of 47 Table A2-9: Main Steam Line Reading at Release I- I .) 1 1-132 1-133 1-134 1-135 Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Xe-137 Xe-138 Kr-83m Kr-85 Kr-85m Kr-87 Kr-88 V ,rQO ,./ /12-Uq-1.3 1E-03 1.53E-03 2.84E-04 4.14E-03 4.26E-02 3.65E+00 6.40E-02 1.16E-01 6.07E-03 2.43E-03 8.82E-03 5.62E-03 1.16E-01 2.28E-02 1.49E-02 4.26E-02 1 1)Q 1 l7W2 2.38E+00 2.03E+01 8.77E-01 6.6 1E+00 2.83E+02 1.27E+02 5.42E+01 9.08E+00 2.60E-0 I 6.38E-02 2.36E-01 1.83E+00 9.40E+04 4.48E+00 1.27E+00 2.84E+00 Iz Inn Al)V./.LiL-U4 9.77E-04 1.47E-03 1.29E-04 3.73E-03 4.25E-02 3.63E+00 6.3 1E-02 1.07E-01 4.22E-04 4.65E-08 4.67E-04 3.85E-03 1.16E-01 1.95E-02 8.62E-03 3.34E-02 1.60E-09 CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds REV. 2 r E N E R C O N for Emergency Action Levels knliece--Everyp,,fcFr.

Every day. Attachment 3 PAGE NO. 30 of 47 DRILL STAMPEDE User Supplied Information DR LRvso7o.0. t DRELL Dautm1ame: 12/1720U 1i524 Uasr Nam: Un VeIt Alm T1huS r 11mfiermt AWwlgo Data lupffh: Graund bwd t mud ioty 13M mi~hr OruundLr Itudmdf ID degrees Clas s$hiy Class: "D -NeuFrnl Uaiteedtnit Veal Rinse: kT Veat Release Rate atert 2044E+)6 uCi t ser Reacts St DateTime: 12/1720131424 Raluse Start 127172013 15:24 EshiaatedRulgase Thonti LOD hmurs Nuock& 3&tre:.Gsph-tery CakuatedNOBLE GAS relese rate: LDE MuCi/sec.NOmIEAS MulisE auI/sec Nadci nuC t sec PAR37CiEATE NMacfi RCisec Kr-l3M: Kr-IS: IKr--L.I&Kr47: Kr-U: Xe--In1LM Xe-133: Xe-I3M: Xe-135M Xe-13T Xe-1Im 253E-+04 1.05E+004 1.74E+005 3.0ff021 3.12Em+t IJ9IE+OW4 1.45E-'0&14E4-003 953ECOXD 266Ef404 1-131: 1-132: 1-133: 1-13: I-135: 1112+02/3.IH+0 6-05B+015 3.120+003 S.12E+0-3 Cs-.LM: Cs-137: Ce/Pr-3M4: Ce-!41: I.2l,40: lI'b-l99: Fm/Rh-lu&Rm-103: Sr/Y-w.Sr-3g: Te-132: zir-95.-1 DEa+l000.105EM10X. 2.1024E10 2.SW4D4 5i3lE-COD 1122E4)0 1352-WDI 3.13Eql01 12/1712013 3:24:46 PM CALC. NO. STPNOCOI3-CALC-002 Radiological Release Thresholds REV. 2 E N E R C O N for Emergency Action Levels v bxceflence-Eveyproject. po veryday, Attachment 3 PAGE NO. 31 of47 DRILL ~STAA-REDE ResultsIfoato DRILL R g DRILL lhf&Mr/IWs1172013 15:24 Uer Nan: Unit Vent Alert Cnuraammznnfl Distauca (miles)05 1.0 2.0 5.0 7-5 10.0 20o Distaxcv (Inits)0.5 1.0 2.0 5.0 7-5 10_0 Distance (sniles)0-5 1.0 2.0 5.0 7-5 10-0 20.0 Ptlmxe Trsrl Tisne ( nuns:Msmntas) 0:02 0.05 0:23 0-34 0:45 1.31 CmLQ Va.ln (Sod=*) 1.032E-00 3.755E-05: 1.00'E-00 3.951E-007 L541E-0Of7 CM BinY 243S005 3.15EE-OD5 737T4-0O?2-441E-007 9.10912-00 la tz Ikse Dow as Im rsion fl al Body 171abrie Z,. 1C~amm 0""0.009 0.0033 0:001 0:0(X)0.000 Isa~rsicn fte Body naoil gus gmass (rein)0.009 0..001 O:0(X)external + anto: 0.016 0.002 0.001 0.000 0.0w0 Imfi a. CUE rut (rusr" 0.137 0-051 0-018 WXM 0.1112 0.001 0.000'IMIE Idize CDE exernsl + internal Muraiu (rem) (rem)0.016 0.006 0.0M2 0.001 0.000 0.000 0.000 0.137 0-051 0.018 0:004 0.002-O.OO2 0.001 0-M0 12317/2013 3 .2428 PM CALC. NO. STPNOC013-CALC-002 for Emergency Action Levels REV. Radiological Release Thresholds Excellence--Every proect. Every doy. Attachment 3 PAGE NO. 32 of 47 STAM,,PEDE Results Information DRILL R-iina 7.0.3.3 9%W82011 Page2of2 DRILL]CalcuLation Comphtedt RESULTS iMe&odofProjecdon: WindVelocity: 132nIr Release Rate: 119E+006 uCVsec STAMPEDE WindDirection: ISO OffMie Dose Projection (rem): 1 mile 2 mile.? S, miles' 10 miles TEDE 0.006 0.01)02 001 CDE 0.051 0).018 0.004 0.001 Projected duration ofreleme:

1.0 hours

A General Emergency Requires a Protective Action Recommendation EVACUATE ZONE(S): I SHELTER LN PLACE ZONE(S): 2 AFFECTED DOWeNWIND SECTORS: R. A, B All Remainine Zones Go Indoors And Monitor EAS Radio Station Based on a Dose Rate Projection of> 3 mreminir (Immersion Ibole Body Noble Gas Gamma) at the Site Boundary (1 Mile) for 15 minutes or longer the Emergency Classification Initiating Condition RAIl (ALERT) has been met.PERFOR\IED I'Y: RIVX[IEWID BY: Kid Lanager.a~dioltgical Director 12117!2013 3:24:44 PM DatT'Time DareTime 12117P-013 3:24-.2S PM CALC. NO. STPNOCOI3-CALC-002Radiological Release Thresholds A E N E R C O N for Emergency Action Levels REV. 2 Excei'lence--Every Everyday. Attachment 3 PAGE NO. 33 of47 DRILL STAMPEDE User Supplied Information Da4fr/Tn: 12'1&2-13 07:54 Cammnts: 9sr Name: Stesminti Site Alt IUT Si Dkwbrfata Input: hndleel ,Mnd'elofltr 132 mil/hr Groumd "ad dfrrn: 130 d&grees tker-select-tedShlYl class StuImtyC is: "D- &bnitwed S)GTnube1pure RIdease: Stea= Adt t5E+000 uC1i/cc StezmT1hwlate: L050 Mlhbr Radacr Shm DalteTme-: Rinms Start DatTimn: FklmatedRdeas Duration: 12/1312f13 06:54 n1mn21on1 07:54 10D hears Nudi& Matrme: Mtle Gas +Icime lon. am mpm-atf of gz& 02%CakrlatedNOBLE GAS rn rate: 1gE+07 wCi/sec NOELECAS IODIME PAYCIITIIUE MaCB& uCi/sec MucUSl umCI/sec Kr-UM: 1r47: Xe-13&Xe-Mml Xe-IM: xe-13at Xe4137-Xe-lit: 1.1,1E-t-(3.43E+00X5 5-7E40004 255NE+004 4.2E-0043 1.0lM-t0D7 3.IZENXIS 123E40+M 127E&101 I3ffiXM 1-131: 31)E+003 1-132: 3.22E+003 1-133: 42824GB 1-134: 41229+002 1-135: L23E+00t Cs-li: Cs-137: ce/r-13'/: Ce-141: 1.a-lAO: 3&t-99: RaIRb-106: Sr/Y-90.Sr-": Te-412: Zr-95: GOJE4WO 0.0aE24000 GOIXJ-000 0.00E+000 O.OaE+COO O.XE+tX0O 0.3m+1000 0311+030 121M 13 7?55:19 AM CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds REV.2 E N E CO N for Emergency Action Levels cxceldence--e'y projwer. day. Attachment 3 PAGE NO. 34 of 47 DRILL STAMPEDE Results Information DP,07033 111 I g laf2 DR IL l "mrzEn 12/1, 20130754 lw Nam: StemI.,m Aat cammmmis 05 1.0 2.0 5.0 75 1D.0 2U.Duuance 0.5 1.0 2.0 5.0 73 1.0 2.0 5.0 7.5 10.0 20.0 mam 0:02&23 0-34 0:45 131 CW]/Q Vain.3.750-406T 1.00_-00v" 5.'D42-001 1.5412407 C D2 DMI 2-43S005 9.11aE-006 7-373E-097 2_441E-0W7 9.10-S-kkzs~uraie frost Rufus Iamerniaa Wfmlel Bmdy noble go gamma 0.011 0o042 0.002 0.000 o.ooo omoo 0.000 anoble em gamma (reim)0.011 0.004 0.002-0.000 0.000 IPAGDos RaisIodine CDE external +.iflerual MUyrei (--Ifb) (--11r)0.019 0.135 0.003 0.017 0001 0.017 0.001 0.(04 0.000 0.000 0.001 0.000 0.000 IME Iodine CDE (lerui+ imainual '&Tria 0.019 0.007 0.001 0.000 0.000 0.135 0.050 0.017 0.004 0.002 0.001 0&0 12A &'2013 754:42 AM CALC. NO. STPNOCOI3-CA LC-002 Radiological Release Thresholds t3 EN ER CO N for Emergency Action Levels REV.&~celence--Everyptoje.r, 'vetydoiy Attachmnent 3 PAGE NO. 35 of 47 DRILL PaTA , Results IoraaionILL DR LLRE%'iion 7.0.3.3 9,29,72011 Page2of2DR L Methodof Projection: STA:MPEDE Offsde Dose Projection (rem): I mile TEDE 0.007 CDE 0_050 l C.acultiom Completed RESULTS Wind Vlocity: 13_2nihr Wind Direction: 180 Release Rate: l.19E+07uCiisec 10 miles 0-001 2 mifles 0.003 0.017 miles 0.001 OI0w Projected dsn-afion of'releae:

1.0 hours

A General Emergency Requires a Protective Action Recommendation EVACUATE ZONE(S): 1 SHELTER IN PLACE ZONE(S): 2 AFFECTED DOWNWIND SECTORS: R, A, B All Remaminin Zones Go Indoors And Monitor EAS Radio Station Based on a Dose Rate Projection of > 3 mremnir (Immersion Whole Body Noble Gas Gamma) at the Site Botmdary (1 Mile) for 15 minutes or longer the Emergency Classification Initiating Condition RAI (ALERT) has been met.PERFORMED BY: 'Name REKIIIMID BY: Rdls'd. nage r:Ramdolo gic al Director 121&92013 7:55:14AM DasThne 121212013 7:54:42, AM CALC. NO. STPNOCOI3-CALC-002 Radiological Release Thresholds REV. 2 E N E R C O N for Emergency Action Levels _EV. 2 Exce ence--ver eptoe. Every day. Attachment 3 PAGE NO. 36 of 47 DRILL STAMPEDE User Supplied Information ,71.033 9/O DRELL Dthfrrie: 12/17/201315:25 commmts User Nam*: UEit Mew Site Ama Iar S- s, mnztk I lh Impxf: Qomn1k-dfl tadikctf 132 nt&r Qonmile~wi dfrsrr 1la &grfees lker-siec-tedStzliiy Class Sthiy (Clas: "D -Neufr." niirmpedklf" Vat 1Rine: Ilit Veat Rlase Rate mteretd 2Z--1007 Reatdr Shutdm Dahtsrne: Rinse Start DatiTume:dRelnseflnraliow Nni& Iftdure: 12117fD13 14-:25 1217/2013 15:25 100 hems Qzpbnntury CakuatedNOBLE GAS rmnse rate: 1!91-7i7u -e NOBLEGAS N Iadi Cv1sac Nadide RCi/sA, MucS& mCilsec 10"-80_K,,r-w9: Xe-.131 Xe-13i XL-135M Xe-137-2527E405 1.fl 7.06E+M]9.0AE.4X15 I:73E+ODS 3.14E-03DX 3.121+0X14 621+-W0 191E'005 1+J-NX14 9.74E+0X1-131: 1-132: 1-133: I-lM:--135: 1-135: 301SE--N-5.12E+004 Cs-1i: Cs-137: Cemr-144: Ce-Idl: Rn/Rh-l06t Sr/Y-w-Sr-": Te-132: Zr-0: 1.05M+01 924E+000 2.GE-+O0 2.94M.3 531rE-03 3.+43o000 i24E,-OD4 25GE4003 1XIG-OO2 1.82Edl)135E+tOD 3.LV3-M 1247T2013 3"2533 PM CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds r E N E R C 0 N for Emergency Action Levels REV.2.-, y; day. Attachment 3 PAGE NO. 37 of47 DRILL STAMPEDE Results Information DRILL Rv=03%M11 Page In!2 DRILL iDak(Tz 12fl.,2D13 1525 w N'ame: Util Vnt Site Ara I Dihnaz (miles)0.2.0 5.0 7-5 10.0 20.0 Diltauc, (Mmiks)1.0 2.00 7.5 IDG 2.0L 5.0 7.5 l.o 20.0 Pinm. "wel TMme CELDQ Valn.@ounmiamos)(soc/a') 0012 2.&,.005 0105 L0372E-00 0:09 375SE-0 0-23 1.01D-006 0-34 5. 7E-007 0:.45 131 1541E-007 2-43dE-M0 9.11CE-00 3.151E-OW 77338-007(3.2458-00 2441E-0 9.109E8-E Immursioa fob Body xale Zu gunoan (r.am~r)0.082 0-032 0-012 0.013 0.002 0-aol ImmersioaInoal Body aRti. zgofmm. (rem)0.012 0:003 0.002 0-001 0ACO0 IPAGDosoeHalo MTE Iodine CME (ron/ic~) (rom/ir)0.160 1364 0.060 0510 0.021 0.176 0.005 0141 0.003 0.021 0.,00 0.014 0.0101 0.005Iodine CDE oxteroul + intern!j MUTSUi 0.160 0.06D 0.021 0.005 0.002 0.0(01 1364 0510 0-176 0.041 0.021 0.014 0M005 1217/2D_13 3:25:21 PS CALC. NO. STPNOCOI3-CAL-C-002 Radiological Release Thresholds CALC. NO.2SPNOC13-CAC-00 EN ER CO N for Emergency Action Levels 1EV.2 Fx,, cP-11enrydojy. Evetyda. Attachment 3 PAGE NO. 38 of47 DRILL Results information 9)2ge2011 Page2of2 Ca_.-intions-Competed RESULTS DRILL leledodofProjectioa: WindVelocity: 132 mi/hr Relee Rate: l.19E-'07 uiQ STA-MPEDE Windflirerdon: 180 Of0ste Dose Projection (rem): I mile 2 mile- Pmils 10 mile!TEDE 0.060 0.021 0.005 0.002 CDE 0.510 0176 0.41 0.014 Projected duration ofrelease:

1.0 houTs

A General Emergency Requires a Protective Action Recommendation EVACUATE ZONE(S): 1 SHELTER IN PLACE ZONE(S): 2 AFFECTED DOWNWTIND SECTORS: R, A, B All Remaining Zones Go Indoors And Monitor EAS Radio Station nec Based on a Site Boundary (1 Mile) Dose Projection > 0.1 rem TEDE andlor 0.5 rem Thyroid CDE the Emergency Classification Initiating Condition RS 1 (SITE AREA EMERGENCY) has been met. I PERFORMED BY-Name REVIEWED BY: Rod Mhnager:Xnadiological Director 12/17/2013 3 25:28 PM Datv'Tfie Dat&eTime 12'17fl013 3:25:21 PIM CALC. NO. STPNOC03-CALC-002 Radiological Release Thresholds REV. 2 , E N E R C O N for Emergency Action Levels Exceilence-Everyptojecr. Everyday, Attachment 3 PAGE NO. 39 of 47 STAMPEDE User Supplied InformatiOn DRILL DRILL DRism7.033 9M01I Dhf/imm: 12f17M2._3 1513 Ceomatm Slr NamE: StEM LI AXea IThaf SumuSed]Oeqmdlewi adiwlocity Ifl nil/h Grmmdie tmdsfdrkn 110 Lgrees Csm StbbmT Ycl : "D -Neutarx mifsklmed SGTWIhRpInre Reies Steam Arftif A45RE+001 uoCt/c StemmlhwRzte: LownlEbr Rector Shut&=e m DateTime: 12/171913 1413 Rase Start DatI/Tume: 12/17/201315:28 hiuntedRleas Dwatori LOG homrs Nndi& MAfhre: NuMo Gas +LItme J&O.. 2S p~rC4t oftmold gas: 01%CakuktINOBL'E GAS ralas rat.: 1102-001 m /sec NOHILEGAS NmUtb& aG/seC IODINE Nudkh o42i/s PARflCELIEM Numcis .0/sec Kr-83M: Kr47: Kr-97-X-13DIM XL-133M XL-lZM: XL-Un: Xe-137-XeL-U38 1.14E+00 3.4.5-+00 S1IE+005 2.55E+005 91E7+005 127E+006 3.SE$XJO6 121E-004 114E4004 1]-131: 1-132: I-lM2 1-133: 1.-135: 323E+0OG 4-9aE400$1-24E4005 Cs-134: Cs-UL7-Ce/Pr-lit: Ce-141: I.e-lO: lb-l(: RR-l03: Sr/Y9t Sr-49: Te-OS: Zir-9: O.O+W00 O.C(E+(O0 O.OA+ODO O.O[E+O000 0.3E+000 0.0M+000 O.XE+O00 0.O4IE-tO0 O.DXI+OOD 12J/7(/013 329:03 PM CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds REV.2 F ,.E E C Nfor Emergency Action Levels Eyctllence-Everypo]im. Evetyday Attachment 3 PAGE NO. 40 of 47 DRILL STAMPEDE Results Informato D-RILL Psgelaf2 DRILL Ilairm 12117.,2013 15:28 Thu Na: StamLnM SkB Arm CUmmEts: mf15..usmiane=hJt I (miles)03 1.0 2.0 5b0 7-5 10-0 20-0 Distance (mniles)0.5 2.0 5.0 75 10-0 2O00 DiSLfJe (Mniks)03 1.0 2.0 5.0 75 20.0 Plume T~rral Tuao 0:05 023 0-34 0-45[31Value 21696-00 1.032E-00 3.45E_-007 1.0.,'1.-0O6 3.9513-42 US4IE-007 CHVQ DflL (cecj')2-436005 9.11C.-006 7373E4-0 2-441E-107 9-109E-.0 I& M .Doe Rw Isa~inn Whle Body 0.111 0-0423 0.015 0.004 O-O2 0.001 01131 Imaniaz Wfole Body oatblh Mgs gamm (rem)0.111 0.042 0.015 0.004 0.002 0.001 0.001 7Ml orterusi +.aturnl 0.1w9 0.072 0.025 0M6 0.002 0.001 LODose Rates Iodine CD]1354 0.506 0.175 0.(041 01Ml 0.013 0.005 5 !E Iodize CDE extrnal + inieoal Tkyraid (cern) (Mm)0.189 1354 0.072 0350 0.025 0.175 0.06 0.041 0003B 0.021 0.002 0-013 0.001 01M05 12117i2..013 23".53 PM CALC: NO. STPNOC013-CALC-002 SRadiological Release Thresholds jE N E R C 0 N for Emergency Action Levels REV. 2 Excelle,,ce--Everyproect., ve.,day. Attachmnent 3 PAGE NO. 41 of 47 STAMNIPEDE Results Information DRILL id~onTO3 3 9!2&'2011 Page2of 2 DRILL Ca-u'donCti ompleted RESULTS Wind Velocit- Wind Dir eion: Igo Release R-te: 1.20E+OOS uCilsec Mlethod of Projec don: STAMPEDE Offsile Dose Projection (rem): Imile TEDE 0.072 CDE 0.506-miles 0.025 0.175miles 0.006 OL041 10miles 0.002 0,013 Projected duration ofreleae: h.O thour A General Emergency Requires a Protective Action Recommendation EVACUATE ZONE(S): 1 SHELTER IN PLACE ZONE(S): 2 AFFECTED DOWNWIND SECTORS: R. A, B All Remaining Zones Go Indoors And Monitor EAS Radio Station Based on a Site Boundary (1 Mile) Dose Projection > 0.1 rem TEDE and/or 0.5 rem Thyroid CDE the Emergency Classification Initiating ConditionRSl (SITE AREA EMERGENCY) has been met.PERFORMED BY: REVIIWED BY: RadiiiLd.tntger.,'.adiological Director I 17T2013 3:29:00 PM Daterime.Daterime 12'17fl013 3-2&:53 PM CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds EN ER CO N for Emergency Action Levels REV. 2 Excelence--Evey projec. Every doy. Attachment 3 PAGE NO. 42 of47 DRILL STAMPEDE User Supplied Information DRML PxW-47.0-33 9z2f01W1DRL Da mdt e : 12d17i2013 15:26 C-nmis ur fNam: Uni 'it (ekmw l=au Sulitdmzt I LbmrtqwdeDatallhhopmt: Gr(mmudlk aMdwlo-ity 131 mit/hr Qound IuL tdfranr 10 &O gree Uer-e ttedStahly Clm Slaby clas: "D -Netufri Itndtme~dUhl Vent Rinse: 11oi Vent eRate terae 2.50E+CM nCi/sec 1Reat Shuttun Dat/lutme: 12/1712113 14:26 Rinse Start Datg/rTi: 12117/2013 15:26 FltinmtedRolmsefDum-io-1-00 hImrs Nucdi Mhlnre: Claphntnqy CauIcuedNOBLE GAS rne rate: LiB'O u -ilsec NOBLEGAS Nudkide m/sK IODMIE Nuduikb RUiSK PAR3fCL.AUf NucL& mCI/sac Ir-tm KY-IS: 1K,--45M Kr M7: lKr-4S: K,,r-It. XTe-lW 2M2E+006 1.0-,M+006 7.0Z2+006 9.03E+_0 1.7313+007 3.1(E+001 3M22E+0X7 6=l+007 191E+al6 1.45E+0X179.6fE-006 1-131: 1-132: 1-133: VIM: 1-L35: 3.1914M 3.0811+00 5.12E-405 Cs-lid: Cs-137: celpr-lU: Ce-141: La-lID: 11,1-99: Nm/Rh-l~dL Sr/y-w: Sr-": Te-132: Zr-95: 1.05E+aD2 6.25E-(X)I 82.1002 2.1E--002 531EOD 3.43EWCC1.25FA-0B9.1CE-003 1.321-001 3.13M-0 121q742013 3:2637 PM CALC. NO. STPNOCOI3-CALC-002 ~ E N E R ~ NRadiological Release Thresholds REV. 2 C 0 N for Emergency Action Levels Y-pr. Attachment 3 PAGE NO. 43 of 47 DRILL STAMPEDE Results Information DRILL %Pagelaf 2 DRILL Daftznrm 12mfl2013 1526 UsNlk ne: CIEMMa, mflur I~aftg Dixtnsa 05 1.0 2.0 5-n 7-5 10.0 20.0 Distuca 05 1.0 2.0 5.0 75 10.?20.0 Diunca (siles)0-5 1.0 20 5.0 7.5 13.0 20.Plus. bad Min.(hecunsmiurths) 0115 0-23 0-45 131.C111V YangJ L032E-005 3.755-1"06 LiME-GOW 5.7DE-077 I1-W1E-007 CHUQ UM'2.43SE-0 9-11tH-GM 31521E-0IM 7372-007 2-441E-00 9.109E-00&easunde DoseRt Isrieu flu! 5e4m 0z79 03322 0,117 C.C29 0.016 0.010 0.00M Immr"ion Wfk.le B-dy nabul gas gmma (rum)0.879 0-332 0.117 0-M9 0.016 0.010 0-03 MEm lodiza CD, exerrsl + itenlau "lyraid (remskr) (nsir)1.5_ 13.64 0.601 509 0210 13,62 0.050 0.411 0.027 0214 0.017 0.135 0.006 OM05 ME bidine C DE cuornu!l + jut.rual ¶Ikyri-d (,rum) (res)1.598 0.601 0210 0.050 0.027 0.017 0.006 13.646 5(3M 1-762 0L411 0214 0.135 0.050 1211..013 3 2 625 YM CALC. NO. STPNOCOI3-CALC-002 Radiological Release Thresholds R EV__2 J E NERC0 N for Emergency Action Levels REV. 2 Fxcelence--Ever/project. Every day. Attachment 3 PAGE NO. 44 of47 STA-MiPEDE Results Information DRILL Rev-ion7Oit3 92Wi3011 Page 2of2 u DRILL C.cultaliorn Complted RESULTS WindVelocity: 132ni/hr Windflirection: 120 Release Rate: 1.19E+-0O uCisec Afethod of Projection: Offsite Doe Projection (rem): I mile 2 mile 5 miles 10 miles TEDE 0.601 0210 0.050 0.017 CIDt 1-62 0.411 0.135 Projected durta-on ofreleme:

1.0 hours

A General Emergency Requires a Protective Action Recommendation EVACUATE ZONE(S): 1, 2 SHELTER IN PLACE ZONE(S); 6. 11 AFFECTED DOWNWIND SECTORS: R, A, B All Remaining Zones Go Indoors And Monitor EAS Radio Station Based on a Site Boundary (1 Mile) Dose Projection > 1 rem TEDE and,/or 5 rem Throid CDE the Emergency Classification Initiating Condition RG1 (GE NERAL EMERGENCY) has been met PERFORMIED BY: Name RfMIEW-ID BY: Kad .iinagerRadiological Director 11q 7/2013 3:26:33 PFM Dateflime Daoe.Time 121 7,2013 326:25 PM CALC. NO. STPNOC013-CALC-002 r Radiological Release Thresholds REV. 2 IAeO E for Emergency Action Levels 3EN.4 F.xce.,flnce-E-Evpy profect. Every day. AttachmIlernt 3 PAGE N O. 45 of 47 DRILL STAMPEDE User Supplied Informaion DRILL DRMDRILLm.-33 f&qI Dhhilme: 12q7,2013 1530 CUMMnRh e-r Namme: SteamLime If-rmo ndll ,tadwabc.ty 1-2 ni/kr Qacnlmkewdi dfrimn I0 yCIlass Stibilty Chus: fl -Nmenr" 3tnihorMdSJGTuhe Raipiur debasc:St 45-M+@02 uCiWc Stem fl=wRate: LONO mllrO Bactar Shut&= Datwflme: 12/17VIh13 14:30 Rmnse Stat BatWklma: 12/17l2013 15:30 BMtantedR1ease Duratiar 1. kourn Nmrh& 1skim:e NoM. Gas +Jmhte lotfm as pwc-rt ofnade gas: C26LNO.LECAS reeae rute: 12 009 nCi/sec NudILEnC/S Nudide IODMLE Nudide MiUMe PARIICLAIE Nurfida aCilsec fl-ISM Kr-IS: Kr4ffM KT497: KY-It Kr-St Xe-l31n&xfL&.12-133k!XLe-IS: Xe-13SS XL-137-12u?-Mt 1.14E+006 3.45E-IXT12SM4U06 925E+006 4.13E-01 127E4UJF7 3.LqE-0417 12M1+005 124F-+00!135E'+005.-131: 1-132: I--1]: 1-L03: 1-134: V1,.5: 3.24E-05 4-QE4005 4214H+00t Csr-134: Cs-137: Ce)Pr-1dA: Ce-141: La-iS0: Mkb-99: MdMR-106: En-l~l: Sr/Y-gI:.Sr-ID: 0.IX4O O.DEE+CX)o 0.04E+C00 OAXE4000 O.D0E+000 ,O(E+OO0 0.UfE-NXO0 0.(XE+000 O.4E+00 Te-132: O.X+CCO0 Zk---: 0.CCE+O0 12A1712013 3:3035 PM CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds F4ENERC 0N for Emergency Action Levels REV. 2 Fxceflence--Everproject. Every day, Attachment 3 PAGE NO. 46 of47 DRILL STAMPEDE Results Information DRILL7.0dskm23&3 t2O11 Pge lo2 DRELL DN rnr 12172013 15:30 Us P&M: Stam.n CGEMMmts: (miles)05 1.0 2.0 5.0 75 103 1.0 2.0 5.0 7_5*(miles)0.5 1.0 2.0 5.0 75 1.10 2CLO Mn Plume Trarel S.m (iunrsrmihetes) 0:02 005 0:23 0-34 0:45 131 CULtI VYang (ser/m/)I.03MM00 3.75.-54-06 5.004-006 3-ME-007 154E-o CHUQ DI1PL 2A43d-005 9.11CE-,W 3.151.E-.D 7.373E-00 3.,845,-007i 2441E-007 9:109.E-I ra. Dose R I ioneniea lde Body 7abl0 zs pummz (0DEm/k)1.109 0L4236 0-153 0.040 0JO22 0.015 Immersion Whlel Body noble jas p tmms (rne)1.109 OL424 0.153 0.040 0JO22 0.015-OM6 LPAG Dos Rates MEh Iodinea CDE extersal+ intiruer. fyroi'(rem/kr) (remir)1.295 0.717 0254 0US 0.034 0022 0.008 13537 5.1!58 1-747 0.211 0.134 0.049 MME ZIodine CDE extrnal + intnrnal Murvid (remn) (rem)1895 0.717 D0254 0.063 0.034 0.022 0.008 13537 5.-1.747 0211 0.134 0.049 CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds REV. 2 EN E R C 0 N toraEmergeny ActionE Levels REV. 2 Fxpene Eey pojet vefydoy Attachiiieiit 3 PAGE NO. 47 of47PEDE Results Information DIRILL ,Rion 7.03.3 9-2&2011 Page2of2 DFRLL Calcultions Completed RESULTS MedthodofProjection: WindVelocity: 132m.,ai Release Rate: 120E+009 uCi'sec STA-MPEDE WindDirection: 180 Oite Dose Projection (rem): 1 mile 2 miles Smiles 10 miles TEDE 0.717 0254 0.063 00o2 CDE 1,747 0.407 0.134 Projected duration ofrelease:

1. hours A General Emergency Requires a Protective Action Recommendation EVACUATE ZONE(S): 1, 2 SHELTER IN PLACE ZONE(S): 6, 11 AFFECTED DOWNWINXD SECTORS: R., A, B All Remaining Zones Go Indoors And Monitor EAS Radio Station Based on a Site Boundary (1 Mile) Dose Projection

> 1 rem TEDE andfor 5 rem Thyroid CDE the Emergency Classification Initiating Condition RGI (GENERAL EMERGENCY) has been met PERFORMED BY: Rn=7D BI. :Name RNlEW1TD BYD Bad ihanager.,Rmdlological Director 12/1712013 3:30&.48 PM Date,/Tfie. DateMime 1217/013 3:3039 NP CALC. NO. STPNOC0 I 3-CALC-004 ID E N E R C O N CALCULATION COVER SHEET REV. 2.xcefl>n fe--ery ptoP G N Every" da PAGE NO. I of 17 Title: Fission Product Barrier Failure for EAL Thresholds Client: South Texas Project Project: STPNOC013 Item Cover Sheet Items 1 Does this calculation contain any open assumptions that require confirmation? (If YES, Identify the assumptions) 2 Does this calculation serve as an "Alternate Calculation"? (If YES. Identify the design verified calculation.) Design Verified Calculation No.3 Does this calculation Supersede -an existing Calculation? (If YES, identify the superseded calculation.) Superseded Calculation No.Scope of Revision: Revision I incorporates the source term provided by STPEGS.Revision 2 removes an incorrect assumption of the RCS density and changes the Safety-Related designation to Non-Safety Related.Revision Impact on Results: For Revision 1, the monitor readings decreased using the updated source term.For Revision 2, the results decreased from 0.448, 137, and 547 R/hr for: the RCS, Fuel Clad, and Containment barriers to 0.134, 40.2, and 383 R/hr respectively. Study Calculation 11 Final Calculation Safety-Related Non-Safety Related (Print Name and Sign)Originator: Caleb Trainor Date: 2/16/2015 Design Reviewer: Chad Cramer Date: 2/17/2015 Approver: Marvin Morris , Date: 2/17/2015 ENERCON CALC. NO. STPNOC013-CALC-004 Eceienrp-Ev ey eery CALCULATION REVISION STATUS SHEET REV. 2 PAGE NO. 2 of 17 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 0 03/04/2014 Initial Issue 1 03/21/2014 Updated the source term 2 02/17/2015 Designate as Non-Safety Related PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION 1-8 2 APPENDIX REVISION STATUS APPENDIX NO. PAGE NO. REVISION NO. APPENDIX NO. PAGE NO. REVISION NO.A 9-17 I CALC. NO. STPNOC013-CALC-004 E N E R C O N Fission Product Barrier Failures for EAL Thresholds REV. 2 Excelen---vry project. Every oy F PAGE NO. 3 of 17 1. PURPOSE AND SCOPE The purpose of this calculation is to determine whether a breach of Fission Product Barriers has occurred based on containment radiation levels. The calculated levels will be used as threshold values in the STP Emergency Action Level (EAL) Technical Basis Document which implements NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors" (Reference 3.1).The failure of the fuel clad barrier, reactor coolant system barrier, and containment barrier will be analyzed individually. Values will be derived for the containment high range radiation monitors RT-8050 and RT-8051 for the three barriers.2.

SUMMARY

OF RESULTS AND CONCLUSIONS The results of this calculation are listed below.Table 2.1: RT-8050 & RT-8051 Response to Fission Product Barrier Failure Failure Monitor Reading (R/hr)Reactor Coolant System Fuel Clad 40.2 Containment 383 Readings on these monitors at or above the values listed indicate a failure of the associated fission product barrier.3. REFERENCES 3.1. NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Rev. 6.3.2. PSAT 3075CF.QA, Dose Calculation Database for Application of Alternate Source Term to LOCA and FHA for South Texas Project Electric Generating Station, Rev. 3.3.3. A41009-00458UB, Unit 1 RCS Volume and Temperature Assumptions for Evaluation of Radiation Sources, Revision 6 3.4. Offsite Dose Calculation Manual, Rev. 17, March 2011.3.5. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 2.3.6. STPEGS Drawing 9C129A81105, Radiation Zones Reactor Containment Building Plan at El. 68'0", Revision 3 3.7. MicroShield 6.20 3.8. Federal Guidance Report No. 11.3.9. STP Procedure 0ERP001-ZV-1N01, Emergency Action Level Classification, Rev. 10, Draft.3.10. ENERCON Services, Inc., MicroShield 6.20 Computer Code Verification, STPNOCO13. 3.11. STPEGS Technical Specifications, Unit I Amendment No. 170, Unit 2 Amendment No. 158 3.12. NOC-AE-07002127, Request for License Amendment Related to Application of the Alternate Source Term Regulatory Guide 1.183 Standard Review Plan 15.0.1 CALC. NO. STPNOC013-CALC-004 10 E N E R C0 N Fission Product Barrier Failures for EAL Thresholds REV.2 Excellric-Fvery projc Fvry doy, F r oe l PAGE NO. 4 of 17 4. ASSUMPTIONS

4.1. Monitor

Location The Containment High-range Area Radiation Monitors (RE-8050 and RE-8051) are located inside the Reactor Containment Building (RCB) close to the outer structural wall. The monitors are mounted approximately 5 feet above the 68'-0" elevation (Reference 3.6). Limitations of the modeling software used (Reference 3.7) requires that the geometry be configured with the monitors placed immediately outside the source volume rather than inside the source volume.This slight difference is assumed to be negligible.

4.2. Dispersion

The entire reactor coolant system (RCS) mass of noble gases are released into containment and evenly dispersed within the Reactor Building volume above the refueling floor (Reference 3.1).However, only ten percent of the iodine is volatilized and released from the RCS (Reference 3.5).It is assumed that all other fission and corrosion products are not released as specified in Reference

3.1 developer

notes.4.3. Buildup Buildup was ignored, as the detector is immersed in the atmosphere. This is conservative, as it produces a lower dose, which provides a lower detector setpoint level.4.4. Activity The activity of the RCS corresponding to the failures of the RCS, fuel clad, and containment barriers are Technical Specification limit, 300 ýiCi/g DEI, and 20% fuel failure respectively as specified in Reference

3. 1. The technical specification limit is I [tCi/g DEI per Reference 3.11, Section 3/4.4.8a.

Reference

3.1 states

that 300 piCi/g DEI is equivalent to 2-5% failed fuel which results in a conservative estimation of 1200 [tCi/g DEI for 20% failed fuel. OENERCON Excelence--very projct. F~vey Fission Product Barrier Failures for EAL Thresholds CALC. NO. STPNOCO13-CALC-004 REV. 2 PAGE NO. 5 of 17 5. DESIGN INPUTS The normal operating coolant inventory (for noble gases and iodine) is taken from Reference 3.12 Table 4.2-20 and is listed in Table 5.1.Table 5.1: Reactor Coolant Inventory Nuclide Activity (iCi/g)Kr831m 3.70E-01 Kr85m 1.50E+00 Kr85 7.60E+00 Kr87 9.80E-01 Kr88 2.80E+00 Kr89 8.40E-02 Xe131m 2.80E+00 Xe133m 4.20E+00 Xe133 2.40E+02 Xe135m 4.OOE-0 I Xe135 7.60E3+00 Xe137 1.60E-01 Xe138 5.80E-0 1 1131 4.25E+01 1132 6..00E+0 1133 7.OOE+01 1134 1.30E+01 1135 1.90E+02 The dose conversion coefficients for iodine are taken from Reference 3.8, Table 2.2 and are listed in Table 5.2.Table 5.2: Dose Conversion Factors Dose Conversion Factors (Sv/Bq)1-131 8.89E-09 1-132 1.03E- 10 1-133 1.58F-09 1-134 3.55E-11 1-135 3.32 E- 10 The containment free volume is taken from Reference 3.2 item 3.2 and is 3.8x10 6 ft 3.The mass of the RCS is 2.658xl 0 8 g per Reference

3.3.6. METHODOLOGY

Equation 6.1 shows the calculation used to find the release concentration of each nuclide. CALC. NO. STPNOCO03-CALC-004 l E N E R C O N Fission Product Barrier Failures for EAL Thresholds REV. 2 of17 Exce en e-F ery pm e Cve Ey doy PAGE NO. 6of1 (XI-13,*DCFI-131]Euto .Release Activity =Ci

• MRCS
• Equation 6.1 Si=13DCF1

/Where: Ci is the concentration of each nuclide in the reactor coolant inventory listed in Table 5.1 ([tCi/g), MRCS is the RCS mass taken from Reference 3.3 (g), The third figure in the equation scales the normal RCS inventory to each barriers specified activity, X1-131 is the concentration of 1-131 specified for each fission product barriers DE1 (lCi/g), DCFi is the dose conversion factors from Reference 3.8 (Sv/Bq), and Ii is the concentration of iodine found in the normal operation reactor coolant inventory (lCi/g).Because the releases are dictated as levels of DEI, dose conversion factors (DCF) must be used to determine release activities. The STP ODCM dictates that the DCF which define DEI are to be taken from Reference

3.8. Using

these conversion factors, the dose related to I ýtCi/g, 300 pCi/g, and 1200 pCi/g is calculated. The dose rate from 1-131 is then compared to the dose rate from the full iodine spectrum in the reactor coolant inventory found in Table 5.1. A ratio between the two dose rates is used to scale the coolant inventory to the dose equivalent specified. This is shown in Equation 6.1 and the tables in section 7.The containment building is modeled in MicroShield as a right circular cylinder with an equivalent volume to the containment free volume. This is filled with the evenly distributed source activity with a detector placed at a location equivalent to actual placement. The MicroShield input and output information for each case can be found in Appendix A.MicroShield is appropriate for use in this calculation and has been approved and documented by ENERCON Services, Inc., MicroShield 6.20 Computer Code Verification, STPNOCO13.

7. Calculations The spreadsheet below contains the calculations described in the methodology section.

CALC. NO. STPNOC013-CALC-004 O ENERCON Exceflenc--Everynprojec. Eveiy doy, Fission Product Barrier Failures for EAL Thresholds PAGE NO. 7 of 17 Table 7.1: Release Inventory for Failed Fission Pr(Nuclide Kr83m Kr85m Kr85 Kr87 Kr88 Kr89 Xel31m Xe133m Xe133 Xe135m Xe135 Xe137 Xe138 1131 1132 1133 1134 1135 Concentration pCi/gm 3.70E-0 I 1.50E+00 7.60E+00 9.80E-0 1 2.80E+00 8.40E-02 2.80E+00 4.20E+00 2.40E+02 4.OOE-01 7.60E+00 1.60E-01 5.80E-01 4.25E+01 6.OOE+01 7.OOE+01 1.30E+01 1.90E+02 Gross Activity ptCi (concentration*mass) 9.83E+07 3.99E+08 2.02E+09 2.60E+08 7.44E+08 2.23E+07 7.44E+08 1.12E+09 6.38E+10 1.06E+08 2.02E+09 4.25E+07 1.54E+08 1.13E+10 1.59E+10 1.86E+ 10 3.46E+09 5.05E+10 Ci @ 1 I Ci/g DEI-131 Ci total=activity* Scaling factor/106 1.57E+00 6.35E+00 3.22E+0 I 4.15E+00 1. 19E+OI 3.56E-O1 1.19E+01 1.78E+01 1.02E+03 1.69E+00 3.22E+01 6.77E-0 1 2.46E+00 1.80E+00 2.54E+00 2.96E+00 5.50E-01 8.04E+00 oduct Barriers Ci @ 300pCi/g DEI-131 Ci total=activity*Scaling factor/10 6 4.70E+02 1.91 E+03 9.65E+03 1.24E+03 3.56E+03 1.07E+02 3.56E+03 5.33E+03_3.05E+05 5.08E+02 9.65E+03 2.03E+02 7.37E+02 5.40E+02 7.62E+02 8.89E+02 1.65E+02 2.41E+03 Ci @ 120OOACi/g DEI-131 Ci total=activity* Scaling factor/10 6 1.88E+03 7.62E+03 3.86E+04 4.98E+03 1.42E+04 4.27E+02 1.42E+04 2.13E+04 1.22E+06 2.03E+03 3 .86E+04 8.13E+02 2.95E+03 2.16E+04 3.05E+04 3.56E+04 6.60E+03 9.65E+04*The scaling factors used in this table are calculated in Tables 7.2 and 7.3**The values in bold have been multiplied by 0.10 to account for the 10% volatilization of iodine during the accident. O ENERCON 0 Excelence-Every project Every doy Y.CALC. NO. STPNOC013-CALC-004 Fission Product Barrier Failures for EAL Thresholds PAGE NO. 8 of 17 Nuclide 1131 1132 1133 1134 1135 Table 7.2:................................ bI e............ A Activity gCi/g 4.25E+0 I 6.00E+0 1 7.00E+0 1 1 .30E+01 1.90E+02 Full Iodine Spectrum DCF Sv/Bq 8.89E-09 1.03E-10 1.58E-09 3.55E-11 3.32E-10 Total =...... ~ ~ ~ ~ ~ ~ ~ ~ ~ ... ....................... tv t

• C Activit y DC F (pCi/g)*(Sv/Bq) 3.78E-07 6.18E-09 1.11 E-07 4.62E-10 6.3 1E-08 5.58E-07 Table 7.3: Scaling Factors Dose Equivalent Activity (ýtCi/g DEI)300 1200 DCF DEA*DCF (Sv/Bq)8.89E-09 8.89E-09 8.89E-09 (paCi/g)*(Sv/Bq) 8.89E-09 2.67E-06 1.07E-05 Scaling Factor (DEA*DCF)

/(sum(Activity*DCF)) 1.59E-02 4.78E+00 1.91E+01 Fission Product Barrier CALC. NO. STPNOC013-CALCO004 r EN ER CO0N Failures for EAL Thresholds REV. 2 Exelene-I'vy polct Fvefyda Appendix A PAGE NO. 9of 17 Case Siumainy of Fuel Clad Barier Page 1 of 3 Micro!'hield 6.20 (05~-MSD-6.2O-1158) Enereron Fe~M 1Re2f1 DO Fil I,-,.Cftecked_Case Tile-- Fwel Ciad BafTi~r Descriptiont: ! P0CGzC Geometrin 8 -CyI djr vIulufi -End1 SIIeI1s 14*1gMt ~ .e3 m (192Ft 11 5 n" Radius .&3c (7E ft)Dot-e P&ýnta A X Y I 50ý.6 CM a 112,64 cmý I j c 2 0it ;,,I 1i93 110 0 in1 j _7ftE _- __ýSN~ew N DimesloaI HAMatai Deni*ty S,"r; 156-s10 C1, Al .012 Source Thput 91 Gu ngj S-a '_w: Ic N of G~p 21 ULAeer *Ile1'gv Ciulo : .015 phoýtmw,3 < 0.015 I lnrudad T,132 2.4Q+09 0 .01 2 36-0973201 4 1-1233 2.65+Q 1.92 11 3 0t55e-X05 .33÷1_13A 5.5000me-131 2.035 Ue + D1, 55ý,629 07, 21 795-00G 1-135~~ CG e + 0400+0 2,9748a+0i

• D I9s0 I 6I 65n 00ýe+CZ00 2.349,75E

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.2339a4+001 Xý- 13 5m ,P I.9 0e +CX0 62530s+010IC .7503es-005 F 6,76f*-o00 Xe-137 6.77D0s-001 7.50C9eI01I7.0-15X-f, 2 .5942 e- 00u1 X I I t2.4C -00+ý" ,0 Y. 1 02e4-" 1 2 54 7 *-Q X 9;42566-001 BuldpTh atrllrdeeic s ouc EbI- ~ N~ Fhaenom Rate Fluevee Rate Exposiare RAte Exposure Rate HeY htot/ mevpvn3/c aC Pmev/cm.laec mR/hr mth#4 Eufu With Uufldap No sul~duP With 11Mu~dp 40.015 2.437ie+12 4 1. 29 1!z F2 1 A 53e , (2 .1 07,E+O I 2461+01 file -MCJProgran%2OFiles%/2 O(x86)/M cro ShieldlExamples%/Casefil es1thflA%1/20 gC L. 2117/2015 Fission Product Barrier CALC. NO. STPNOCO03-CALC-004 0 E N E R C O N Failures for EAL Thresholds REV. 2 Fxcellence-EFveyprojet EveFy yd'y Appendix A PAGE NO. 10 of 17 Case Summary of Fuel Clad Barrier Page 2 of 3 0-43 0.6 0.1 1-0 0.5 20.30.4A TOWS 1.880e+13 2.13 le + 1 1. 804e +11 2-370e+11 2.069e + I 2.644e + 11 J , SOe+1I 2.605e ,10 5. 89B + 03 3.841e+23 I.03?L+00 3.852~e-02 3.232e,03 3. 144 &-02 I. G4 3.e-03 1. 419t -03 L. 459e-03 2. 604e+03 2. 930e- 03 5.755e+0U3

1. 0 1 E+04 1. 1496+3&-C 4.860e+04 S. 1 45e03 2,3130e+04 1, 70 5- 00 5,761e+02 4.394e +03 3 .9686+02 1 261e+03 1 .66'e+(13 1.65r0e+03 2.911e+03 3.17.4e-03 6.156+e03 1.0716+04 1.198e+03 6.55'3e+04 6-.344e-01 5.964e-01, 2-032&+r0O 2-785e+C-D 2,S817e6,0O 4-9526+C0 5-309'e+00 1.sm.e+c01 4.334".2 8.072e+01 3.640ei-01 7.755e+00 7.52 7e-01 2.45ie -DG 3.271e+00 5.548e+0,3 5.s551+00 1.041e+01 1 .857e+01 1.626e+00 4.483e.02 1.8816+02 f~ie:///CJlProgram%2OFiles%20(x86)fMcaoShield'ExamplesICaseFiles/HTh1/1%20-)iCi..

2/17/2015 Fission Product Barrier CALC. NO. STPNOC013-CALC-004 0 E N E R C O N Failures for EAL Thresholds REV. 2 Exc,,nre-E'project, ,vetydoy Appendix A PAGE NO. 11 of 17 Case Summary of Fuel Clad Barrier Page 3 of 3 to., z file..f//C:IProgramlc2OFiles%2o-O(x86)/MfiaoShieldf-xamples/CaseFiles/HTMEl/%2OgiCL.- 2/17,'201 5 Fission Product Barrier CALC. NO. STPNOCO13-CALC-004 0 E ENE R C O N Failures for EAL Thresholds REV. 2 nellec-Fry-roj-Evef day. Appendix A PAGE NO. 12 of 17 Case Sunmmary of RCS Barrier MicroShield 6.20 (05-MSD-6.20-1158) Enercon Page 1 of 3 pa'ge DOS File Run Date Ruan Tlme Duration-t300 WC, DEI.mti6~Fcbruary 17, 01 2ýS3A45 PM M 0,:0:Do0 File Rer Datat By Che-ked (As TfflIei R.Cs Barrier Descriptiom 300 PCI/cc 8 ffeo -Cy!i~ra8r Vdolume -End SýlIelds SI~urce DlnterltonE$ Height 5.9c+3 cm Radius 2.3e+3cm f 192 ft 11 .5 in)ý7sfrt)A*1 x y z 601.6 CmI 582.64 cm 1143 cm 20 lt 4)[)in 11;3 FT00 In 37 It 60 ýShields Shield N Ditmensimi materila souice 9.66+10 CM 3 Air Ail Gap AXT Source Input -Groupinlg Kettbod -Standard Lattices Number of GPMups 25 Lower Ewgy Cutoff, 0.015 pwunts < 0.015 " adalded Ullrary : rv 0.00122 0.00122 Nucitde I- 13t 1-132 1-133 1-134 Xe-137 Xe- 133 5.40001e-a-02 47.600r8+002 1.65008+003ý

1. 91008+0 D 3 1,:40rt!-M3 3.508+0 1,07008a+002

3. koc0e+QC`3.05008+e'005 UC30e03 9 f-S008+003 5 r08008+02 2 0300e+ý002 7 370Ce-002 buecuerel 1.9980e+(013 2,8194e+013

ý332;938+013 7 57Ce+ G13 35705se+016 L1 72le+C141 7261)2e,313-S.5926e-003i 7,8918e-003 9.2071e-c()] 1 7088e-C03 2.496r08-002 4 8676e-C003 9'9942e-002 1978 18-Q02 3 687rje-002 3.6870e-002 3 .1588 e+ý000 5.5201e-002 99942e-GO2 S2612ii-003 2.10248 003 7 532;98-003 2.0693e.002 2.920080092 3.4ý0668002 c.3227e.+001003 7,3-'19 18+ DO2 4 7518ý,+ 02 1-3612t+01)3 4,1002e+101. 1.M42e8-003 1.1687t-+105 2,04248+003ý 3.69788+00C3 1.9486"8002 7.77898+ý0011 2.8242e.002 BLIldup : The maaittLV refaelele IS -SoUfce IneoM" ,rameteis 20 Radial COecumnlerfIlliaI 1t 10 Enetgy 04eV 0. .......015 ACtivYty Flpulca Rate Phorara/ac l~eV/O~n2/** 7.436e-14 ~AI &* 8+9 4 Fterits Fluence Rate Nev/cmli/sec With Buildup 4.34le-04 Exposure Rate mR/hr No Buildup 3.311e+03 Exposure Rate mR/hr With Buildup 3.726e.+03 ffile:W/C /rogram%,2OFilesl20(x86)!NlicroShieldVExampkesiCaseFiles/HTMLJ3O00%2OpiC. 2117"2015 Fission Product Barrier CALC. NO. STPNOC013-CALC-004 0 E N E R C O N Failures for EAL Thresholds REV. 2 F vey da Appendix A PAGE NO. 13 of 17 Case Summmry of RCS Bamer Page 2 of 3 0.03 0.08 0-.15 020 L3o 4.0 1.5 5.622e+ IS 4. 142e -15 2.598e -11 6 .3 10e +13 3.8 7Be+ 14 2.448e. 13 5.954L 13 6. 399 e, 1-3 5.409e 13 7. 104e +13 6.200e 13 8.070e 13 1.ý047e+14 7.787e- 12-1,773e ,11 1.149e+16 I .539e +06 3.ft96+06 3.0O886+,-G

1. 158e,+C5 9.683e+05 9.437e+04 3. 123p, 0, 4.2F 16+05 4. 375f,-,05 7,803e,+G5 8,6314+05ý_

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211712015 I Fission Product Barrier CALC. NO. STPNOC013-CALC-004 E N E R C O N Failures for EAL Thresholds REV. 2 E Everyday. Appendix A PAGE NO. 14 of 17 Case Summary of RCS Barrier Page 3 of 3 z ffiecI//CilPrograml2OFilesl2O(xS6AicroShield/E-xamplesiCaseFile~fl1Tfl30OO%2OjiC. 2/1Ti201 5 Fission Product Barrier CALC. NO. STPNOC013-CALC-004 E N E R C O N Failures for EAL Thresholds REV. 2 Ecelnce-FEvrypret Evedy Appendix A PAGE NO. 15 of 17 Case Summary of Contiamnent Barrier Page 1 of 3 MicroShield 6.20 (05-MSD-6.20-1158) Enercon Page DOSFile Run Date ,Run Time Duration~1200 WCI DEL.ms6 February 17, 2015 2,41 11 PMý.0 -0~0 0 FI'le RA" Checked CA"O Tltleý Coannment R5~fier Degcrlptjoe 1200 pci/cc 6oojetiy. 8 CylInder Volumne -End hed souroi, Dimenst~nsý Radius 2 3e+3 crrn J'7ft)voý,e Points A X Y 61019.6 cmi 5SI3264 cm, 1143 cm 20 ft 0- In 193 TtO0.0In 37 R 6.0iW Shield N sou ce AC;Ap slhfeldt Dimension~ 9 66ce+ 10 cfr S"I'M Input. GroUF~ng Method -Standard radices Mumber of Groups 25 Lower Eney Cu~toff~ 0.01s Photýn S < 0.0 15 -Ind itJed Lib~w,, ý Grove Nucilde 1-131 1-132 1-133 1-134 1-135 Kr F83mi Kr- 3 Kr-S 7 Kr4S8 Kr 89 Xe-131w Xe-133 Xe-133mu Xe-135 Xe-135m Xe-137 2. 1600e41304 3.rO5LK'e+0C,1 3.58E. +004-rQ, 616000e+003 1388(0C*003 7.62OCý" 003 lAaK0e+ 0G4 4.2700e+002 1,4200,e-004 1 .2200e. Out 2.0130ce -+00 8. 1 300e+002 2.950r~ep00 7.9923e+014 1, 28q5e+015 2.442DC-014 3 'i705,4.015 6.95(H05+013 I 4282te+GIS 2,519'4e.s014 !.842t-e+C14 5S40+1 L57995+,113 5.25405+014 4_5140ýe+015 788310etf3!4 1.4282e+L15 7.51 1t0+013 3.D0051a013 1,0915e+-014 3.2370ie-V01 6.8354e-0112 9,99421e-00 1.94 71e 002 3,9977,ý-001 7.869 1Se -CC2 S. 1576.e-cl02 1 .47,_,e-0)1 4 4223.e-003 1 .4706-ke-W 2.20505t-00 3 .99'7e.-C'M

2. 1024a-r002 8.4.,2005e,-V,3 3.0552e-Cr02 Material Density Air 0.100 122 1. 1679li-O0 2.520015+003 1,9083e- 003 5.44114e5+003 4.F750.54.05 5.1621.5-03 1.4791e-QC4 7.77859e+002

3. 11541w,002

1. 13 04 C -00ý3 Buildup ;The matefia reference Is -Source lnl,9 raitio~n Parameters, Radlat 20 YCIrcutetisI 1 Enearg y Mev.....ACMKVit Flue"Ce Rate phafne/ve "v/CrI/e N2o Run-up 29 1e+ 1ý S47e.+i05 Flaence Rtate MeNF/Cm./eer wnh 74 le+0 Exposuere Rate MR/fhr no But*1d 9 p 1. 327e+-A Expojure Rate ma/lw With Buldup 1 443e+04 fiaei//lCirograml,2OFIks%2,?O(x8)MicroShielVExamples/CaseFiks/1TM1J1200%2OL..ý 1217/2015 Fission Product Barrier CALC. NO. STPNOC013-CALC-004 SE N~ E R N Failures for EAL Thresholds REV. 2 Excellence-Every prqect. Evy day. Appendix A PAGE NO. 16 of 17 Case Summary of Contianment Barrier Page 2 of 3 0.03 2.254e16 6.172e,- 06 '9,767e-06 O.117e+04 9.679e-04 0.6 1 -6f$9e 16 ..1.5414e 075f+4 .5e0 0.1 1,03ý6q12 1.232--03

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1. 358e, 07 i34+7 2-651e+04 3.1)53e- D4 0.C8 1. 98 5e -I 2, 11e+07 2,443e +D7 4.14144-04 4.1,46e +04 1.0 2.769ea-07 3.053+07 5 105e+04 5.627e.04 15 2.2768e+15 4.8684+07 5 235e+07 8.1544+04

.0e4 2.0 8.228e+14 2.377e+07 2.518e÷07 3.89et-04 310 3.124e+13 1 .377e+06 1. 437e+06 1.869e+03 1 950e+03 4.0 7..75......... e+1 ..2e...4 4,34F..04 5 19a,-+01 5.3 +o01 Totals S.Sg84+16 1,820e+08 2.156'e+08 3.531*4.05 4.721e+05 file:/IIC~Ilrograml2OFilesl2Ox86)/MicroShiieldExamplesICaseFiles/HTML/1200%2Oiý.. 2/17/2015 Fission Product Barrier CALC. NO. STPNOC013-CALC-004 E N E R C O N Failures for EAL Thresholds REV. 2,yroiect veryday Appendix A PAGE NO. 17 of 17 Case Summary of Contianment Barrier Page 3 of 3 K.file-./I/C:Program%/ý2OFiles*/20(x86),McroShield/Eamples!CaseFilesIHIMJ 1200%/204... 2117,2015}}