ML15068A045
| 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 CompanySouth Texas Prolect Electric Generating Station PO. lat 289 Wadsworth. Teas 77483 -vV/-- --February 26, 2015NOC-AE-1500322610 CFR 50.90File No. G25U. S. Nuclear Regulatory CommissionAttention: Document Control DeskWashington, DC 20555-0001South Texas ProjectUnits I and 2Docket Nos. STN 50-498, STN 50-499Response to Request for Additional Information -South Texas Project (STP), Units 1 and 2 License Amendment Request forEmergency Action Level Scheme Change (TACs MD4195 and MF4196)References:1. Letter; G. T. Powell to USNRC Document Control Desk; "License Amendment Requestfor Revision to Unit 1 and Unit 2 Emergency Action Levels;" NOC-AE-14003087; datedMay 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 (TACsMD4195 and MF4196);" dated December 18, 2014 (ML14352A1 80)3. Letter; A. Capristo to USNRC Document Control Desk; "Response to Request forAdditional Information -South Texas Project (STP), Units 1 and 2 License AmendmentRequest for Emergency Action Level Scheme Change (TACs MD4195 and MF4196);"NOC-AE-15003214; dated February 11, 2015By Reference 1, STP Nuclear Operating Company (STPNOC) requested approval of a LicenseAmendment Request for revision to Unit 1 and 2 Emergency Action Levels. By Reference 2, theNRC staff requested additional information (RAI) to complete its review. STPNOC responded tothe RAI in Reference 3 with the exception of RAI-04. STPNOC's response to RAI-04 inReference 2 is provided in Attachment 1 to this letter. Based on a phone call with NRC stafffollowing transmittal of Reference 3, STPNOC is also including in Attachment 1 revisedresponses to RAI-01, RAI-06 and RAI-12.In Reference 3, STPNOC noted that several discrepancies were found in the calculation thatwas performed to determine EAL threshold values for Abnormal Rad Levels. In the process ofresolving the discrepancy, a review was conducted of calculations that were performed tosupport the proposed revised EALs. As a result, six calculations were revised; two of the revisedcalculations affect STPNOC's response to RAI-04 and RAI-10. The revised response to RAI-10is included in Attachment 1. Revision bars in Attachment 1 indicate changes to the RAIresponse provided in Reference 3 for RAI-01, RAI-04, RAI-06, RAI-10, and RAI-12.STI: 34060108AYqo NOC-AE-1 5003226Page 2 of 3A clean copy and a redline markup of the STPEGS Emergency Action Level Technical BasesDocument is included in Attachments 2 and 3, respectively. Attachment 4 provides revisions tothe STPEGS Emergency Action Level Deviation, Difference and Justification Matrix provided inReference 1. Attachment 5 contains revisions to two calculations that are being provided assupporting documents.The No Significant Hazards Consideration determination provided in Reference I is not alteredby 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 CapristoExecutive Vice PresidentChief Administrative OfficeramrAttachment:1. Response to Request for Additional Information -South Texas Project (STP), Units 1 and 2License Amendment Request for Emergency Action Level Scheme Change2. STPEGS Emergency Action Level Technical Bases Document -clean copy3. STPEGS Emergency Action Level Technical Bases Document -redline markup4. STPEGS Emergency Action Level Deviation, Difference and Justification Matrix -revisionsonly5. Supporting documents NOC-AE-15003226Page 3 of 3cc:(paper copy)(electronic copy)Regional Administrator, Region IVU.S. Nuclear Regulatory Commission1600 East Lamar BoulevardArlington, TX 76011-4511Lisa M. RegnerSenior Project ManagerU.S. Nuclear Regulatory CommissionOne White Flint North (MS 8 G9A)11555 Rockville PikeRockville, MD 20852NRC Resident InspectorU. S. Nuclear Regulatory CommissionP. 0. Box 289, Mail Code: MN1 16Wadsworth, TX 77483Morgan, Lewis & Bockius LLPSteve FrantzU.S. Nuclear Regulatory CommissionLisa M. RegnerNRG South Texas LPJohn RaganChris O'HaraJim von SuskilCPS EnergqyKevin PolioCris EugsterL. D. BlaylockCrain Caton & James, P.C.Peter NemethCity of AustinCheryl MeleJohn WesterTexas Dept. of State Health ServicesRichard A. RatliffRobert Free
Attachment
IResponse to Request for Additional Information -South Texas Project(STP), Units 1 and 2 License Amendment Request for Emergency ActionLevel Scheme Change
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1NOC-AE-15003226Page 1 of 5REQUEST FOR ADDITIONAL INFORMATIONSOUTH TEXAS PROJECT, UNITS 1 AND 2LICENSE AMENDMENT REQUEST FOREMERGENCY ACTION LEVEL SCHEME CHANGEDOCKET NUMBERS 50-498 AND 499The NRC staff requires the following additional information to complete its review of the request:RAI-01Because the information in the basis document can affect emergency classification decisionmaking, NEI 99-01, Revision 6, Section 4.6 contains an expectation that the basis document willbe 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 toensure 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 toensure that changes to Abnormal and Emergency Operating Procedures are screened todetermine if an evaluation pursuant to 10 CFR 50.54(q) is required. This RAI is intendedto ensure similar controls are in place for the STP EAL Basis Document.STPNOC RESPONSE TO RAI-01Administrative controls that address evaluating changes to the STP EAL Technical Bases inaccordance 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) asdirected by OPGP05-ZV-0010, Emergency Plan Change.
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1NOC-AE-15003226Page 2 of 5RAI-04The 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 thediscrepancy in the basis section.STPNOC RESPONSE TO RAI-04STPNOC 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 atSTP. STP does not have Main Steam Line radiation monitor (RT-8046 thru 8049) RU1 and RA1values in the current EAL scheme for comparison. The proposed STP RU1 thru RG1 EALswere compared to another US PWR (Comanche Peak) and were found to be similar. STPNOChas not identified any discrepancies.The STP Radiation Monitoring system is a distributed microprocessor based system consistingof radiation detectors, an associated microprocessor control panel (RM-80), radiation monitoringcomputers (RM-1 1), and digital display modules (RM-23) for the Process and EffluentsRadiation 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 from20 pci/sec to 2E+16 pci/sec and the Main Steam Line radiation monitors have a range from1.4E-02 pci/cc to 1.4E+06 pci/cc.The hand calculated and STAMPEDE (software) calculated values for the RU1, RA1, RS1 andRG1 are contained in Calculation No. STPNOC013-CALC-002 Revision 2, Table 2.1. Table 2.1appears 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 belowshowing the results from all the calculations. The minor difference is due to STAMPEDE usingdecay factors over a one hour period after shutdown. This also accounts for the change in thelimiting dose being TEDE in the hand calculation and Thyroid CDE in the STAMPEDEcalculations. The accuracy of the hand calculation is considered sufficient and recommendedfor 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 arethe unadjusted hand calculated values found in Calculation STPNOCO13-CALC-002 Rev. 2Table 2.1. The Unit Vent RA1 value (1.50E+06 pci/sec) was derived by conservatively roundingthe 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 of4.03E+00 pci/cc.The relationship of site boundary doses between the Alert and the Unusual Event isapproximately a factor of fifty. The gaps between the Unusual Event and the Alert are expectedto be different from the gaps between the Alert and higher classifications based on theunderlying assumptions in NEI 99-01.Radon response at STP is expected to be no more than 1% of the UE level based on operatingexperience and engineering judgment.
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1NOC-AE-15003226Page 3 of 5RAI-06The proposed EAL RA3.2 includes a number of plant areas for all operating modes. Pleaseverify the plant areas identified for EAL RA3.2 reflect only those areas required for normal plantoperations, cooldown, or shutdown, and that access to these areas is required, i.e., cannot beoperated remotely. Please provide evidence of this verification, or revise as necessary tosupport accurate and timely assessment. In addition, consider adding operating modespecificity to the listed areas if applicable.STPNOC RESPONSE TO RAI-06STPNOC has revised EAL RA3.2 to include a listing of plant areas that require access fornormal plant operations, cooldown, or shutdown and components in these areas cannot beremotely operated. Additionally, modes of applicability have been included for each area asfollows:(2) An UNPLANNED event results in radiation levels that prohibit or impede access toANY of the areas listed in Table H3/R2.TABLEEH3/R2: "Pant Areas Requiring AccessRCB RHR Heat Exchanger Rooms0o MAB 51 ft Room 335EAB Roof, MCC 1G8, 4.16KV Switchgear Roomsu E00 "n EAB 4.16KV Switchgear RoomsEAL Selection BasisThe 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, Normaloperations 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 alist of plant rooms or areas with entry-related mode applicability that contain equipment whichrequire a manual/local action necessary following entry into hot shutdown (establish ResidualHeat Removal shutdown cooling, disable operation of charging and ECCS equipment, and limitdilution pathways) and subsequent entry into cold shutdown (disable operation of ECCSequipment). After achieving cold shutdown it is assumed that the plant will be maintained in acold shutdown condition.
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1NOC-AE-15003226Page 4 of 5RAI-10The 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 asthe result of an RCS leak would preclude the use of containment radiation monitors (RT-8050and RT-8051) for approximately 40 minutes, and a secondary system break would preclude theuse 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 forEAL determination, or justify how this information will consistently be used by EALdecision-makers.b. Please explain why these limitations were not included for Containment Barrier PotentialLoss threshold 3.A.1, or revise accordingly.c. Please explain why these limitations were not included for Fuel Clad Barrier Lossthreshold 3.A.1, or revise accordingly.STPNOC RESPONSE TO RAI-10STPNOC has removed containment radiation monitors (RT-8050 and RT-8051) from the RCSBarrier Category 3, RCS Activity/Containment Radiation table due to a reduction in the setpointvalue based on calculation STPNOC013-CALC-004, Revision 2. CalculationSTPNOC013-CALC-004 was revised in February 2015 and would have lowered the RT-8050and 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 effectof 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 canfulfill the function of RT-8050 and RT-8051.STPNOC has revised the bases for the Containment Barrier Potential Loss threshold 3.A. 1 andthe Fuel Clad Barrier Loss threshold 3.A.1 to state that temperature induced current (TIC) is nota limitation for these events.Temperature induced current (TIC) limitations are not applicable to theContainment Barrier Potential Loss threshold 3.A.1 (Fuel Clad BarrierLoss threshold 3.A. 1) because the expected radiation dose for this eventoverwhelms the TIC effect. This is discussed in the 1OCFR50.59evaluation 04-8245-60 associated with DCP 04-8245-33.
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1NOC-AE-15003226Page 5 of 5RAI-12The proposed EAL HA5 appears to cover a wide range of rooms or areas during all modes ofoperation.a. Please verify the plant areas identified for EAL RA3.2 reflect only those areas requiredfor normal plant operations, cooldown, or shutdown, and that access to these areas isrequired, i.e., cannot be operated remotely. Please provide evidence of this verification,or revise as necessary to support accurate and timely assessment. In addition, consideradding 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 plantarea where access is needed to support normal plant operations, cooldown, orshutdown.STPNOC RESPONSE TO RAI-12STPNOC has revised EAL HA5.1a to include the control room as a plant area where access isneeded to support normal plant operations, cooldown, or shutdown.1 a. Release of a toxic, corrosive, asphyxiant or flammable gas into the Control Room orany of the plant rooms or areas listed in Table H3/R2:ANDb. Entry into the room or area is prohibited or impeded._ TABLE H3/R2: Plant Areas Re6uiring AccessRCB RHR Heat Exchanger Rooms0o 0 MAB 51 ft Room 335EAB Roof, MCC 1G8, 4.16KV Switchgear RoomsLu0o Ln EAB 4.16KV Switchgear RoomsEAL 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, Normaloperations 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 alist of plant rooms or areas with entry-related mode applicability that contain equipment whichrequire a manual/local action necessary following entry into hot shutdown (establish ResidualHeat Removal shutdown cooling, disable operation of charging and ECCS equipment, and limitdilution pathways) and subsequent entry into cold shutdown (disable operation of ECCSequipment). After achieving cold shutdown it is assumed that the plant will be maintained in acold shutdown condition.
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2STPEGS Emergency Action Level Technical Bases Document -clean copy STPEGS Emergency Action LevelTechnical Bases Document Rev. 0NEI 99-01 Rev. 6 ImplementationFebruary 2015NOTE: Changes to this document require a review under 10CFR50.54 (q) as directed by OPGP05-ZV-O010,Emergency Plan Change. TABLE OF CONTENTS1 DEVELOPMENT OF EMERGENCY ACTION LEVELS ........ Error! Bookmark not defined.1.1 REGULATORY BACKGROUND ................................................................................................ 11.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) .......................................... 11.3 N R C O R D E R E A -12-05 1 ................................................................................................................... 22 KEY TERMINOLOGY ......................................................................................................... 32.1 EMERGENCY CLASSIFICATION LEVEL (ECL) ...................................................................... 32.2 IN ITIA TIN G CO ND IT IO N (IC) ................................................................................................... 52.3 EM ERGEN CY A CTION LEVEL (EAL) ........................................................................................... 52.4 FISSION PRODUCT BARRIER THRESHOLD ............................................................................... 53 DESIGN OF THE STPEGS EMERGENCY CLASSIFICATION SCHEME ........................... 73.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS) ............................... 73.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS ................. 103.3 STPEGS DESIGN CONSIDERATIONS ..................................................................................... 103.4 ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION ............................... 113.5 IC AND EAL M ODE APPLICABILITY ..................................................................................... 114 STPEGS SCHEME DEVELOPMENT ............................................................................ 134.1 GENERAL DEVELOPMENT PROCESS ................................................................................... 134.2 CRITICA L CH A RA CTERISTICS ............................................................................................... 134.3 INSTRUMENTATION USED FOR EALS ................................................................................ 134.4 REFERENCES TO STPEGS AOPS AND EOPS ........................................................................ 145 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ........................................ 145.1 GENERA L CON SIDERA TION S ............................................................................................... 145.2 CLASSIFICATION M ETHODOLOGY ....................................................................................... 155.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS ........................................ 155.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION ............................. 165.5 CLASSIFICATION OF IMMINENT CONDITIONS ................................................................. 165.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING ............. 165.7 CLASSIFICATION OF SHORT-LIVED EVENTS .................................................................... 175.8 CLASSIFICATION OF TRANSIENT CONDITIONS ................................................................ 175.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION ............... 185.10 RETRACTION OF AN EMERGENCY DECLARATION ....................................................... 186 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/EALS ......................... 197 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS ...................... 408 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS .............. 639 FISSION PRODUCT BARRIER ICS/EALS .................................................................. 6610 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ........... 8511 SYSTEM MALFUNCTION ICS/EALS ........................................................................... 111APPENDIX A -ACRONYMS AND ABBREVIATIONS ....................................................... 140APPENDIX B -DEFINITIONS .......................................................................................... 142 THIS PAGE IS LEFT INTENTIONALLY BLANK 1 DEVELOPMENT OF EMERGENCY ACTION LEVELS1.1 REGULATORY BACKGROUNDTitle 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 anemergency classification scheme. A review of the relevant sections listed below will aid the reader inunderstanding 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 OrganizationThe above regulations are supplemented by various regulatory guidance documents. Three documents ofparticular relevance to NEI 99-01 are:* NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency ResponsePlans 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 § 50.73Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactor1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)South Texas Project Electrical Generating Station (STP or STPEGS) is locating an ISFSI approximately 450 feetwest of the Unit 2 Reactor Building. The STP ISFSI will be within the site Protected Area and is scheduled to beoperational in 2016.Selected guidance in NEI 99-01 is applicable to the STPEGS emergency plan to fulfill the requirements of 10CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistentwith the requirements of 10 CFR § 50 and the guidance in NUREG 0654/FEMA-REP-1. The initiatingconditions germane to a 10 CFR § 72.32 emergency plan (as described in NUREG-1567) are subsumed withinthe 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 thespectrum of credible natural and man-made events included within the scope of the STPEGS ISFSI design. Inaddition, appropriate aspects of IC HU1 and IC HAl address a HOSTILE ACTION directed against the STPEGSISFSI.II P ag e The analysis of potential onsite and offsite consequences of accidental releases associated with the operation ofan ISFSI is contained in NUREG- 1140, A Regulatory Analysis on Eme-ge/ic)' Preparedness for Fuel Cycle andOther Radioactive MIaterial Licensees. NUREG- 1140 concluded that the postulated worst-case accidentinvolving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that themaximum offsite dose to a member of the public due to an accidental release of radioactive materials would notexceed 1 rem Effective Dose Equivalent.Regarding the above information, the expectations for an offsite response to an ALERT classified tunder a 10CFR § 72.32 emergency plan are generally consistent with those for an UNUSUAL EVENT in a 10 CFR § 50.47emergency 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 § 50.47emergency plan (e.g., no emergency technical support function).1.3 NRC ORDER EA-12-051The Fukushima Daiichi accident of March 11, 2012, was the result of a tsunami that exceeded the plant's designbasis and flooded the site's emergency electrical power supplies and distribution systems. This caused anextended loss of power that severely compromised the key safety functions of core cooling and containmentintegrity, and ultimately led to core damage in three reactors. While the loss of power also impaired the spentfuel pool cooling function, sufficient water inventory was maintained in the pools to preclude fuel damage fromthe loss of cooling.Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessaryto ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR50.109(a)(4)(ii). Among them was to provide each spent fuel pool with reliable level instrumentation tosignificantly enhance the ability of key decision-makers to allocate resources effectively following a beyonddesign basis event. To this end, the NRC issued Order EA-12-051. Issuance of Order to Modio, Licenses withRegard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all US nuclear plants with anoperating 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 inassociated spent fuel storage pools capable of supporting identification of the following pool water levelconditions by trained personnel: (1) level that is adequate to support operation of the normal fuel pool coolingsystem, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuelpool operating deck, and (3) level where fuel remains covered and actions to implement make-up water additionshould 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 theused 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 connectioncapability.NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regardto 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 levelinstrumentation associated with NRC Order EA- 12-05 1. These EALs are included within existing IC RA2, andnew ICs RS2 and RG2. These EALs will be implemented when the enhanced spent fuel pool levelinstrumentation is available for use.2 KEY TERMINOLOGY USEDThere are several key terms that appear throughout the EAL methodology. These terms are introduced in thissection to support understanding of subsequent material. As an aid to the reader, the following table is providedas an overview to illustrate the relationship of the terms to each other.EMERGENCY CLASSIFICATION LEVELUNUSUAL EVENT ALERT SAE GE4, 40 +Initiating Condition Initiating Condition Initiating Condition Initiating Condition4, + 4' +"Emergency Action Emergency Action Emergency Action Emergency ActionLevel (1) Level (1) Level (1) Level (1)" Operating Mode -Operating Mode
- Operating Mode
- Operating ModeApplicability Applicability Applicability Applicability" Notes -Notes -Notes
- Notes" Basis
- Basis -Basis
- Basis(1) -When making an emergency classification, the Emergency Director must consider all informationhaving a bearing on the proper assessment of an Initiating Condition. This includes the EmergencyAction Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basisinformation. In the Recognition Category F matrices, EALs are referred to as Fission Product BarrierThresholds; 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 onsiteand 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 plantor indicate a security threat to facility protection has been initiated. No releases of radioactive material requiringoffsite 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 eventinformation and decision-making.2.1.2 ALERTEvents are in progress or have occurred which involve an actual or potential substantial degradation of the levelof safety of the plant or a security event that involves probable life threatening, risk to site personnel or damage tosite equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of theEPA PAG exposure levels.Purpose: The purpose of this classification is to assure that emergency personnel are readily available to respondif the situation becomes more serious or to perform confirmatory radiation monitoring if required, and provideoffsite 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 neededfor protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) towardsite 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 levelswhich exceed EPA PAG exposure levels beyond the site boundary.Purpose: The purpose of the SITE AREA EMERGENCY declaration is to assure that emergency responsecenters are staffed, to assure that monitoring teams are dispatched, to assure that personnel required forevacuation of near-site areas are at duty stations if the situation becomes more serious, to provide consultationwith 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 ormelting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss ofphysical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsitefor more than the immediate site area.Purpose: The purpose of the GENERAL EMERGENCY declaration is to initiate predetermined protectiveactions for the public, to provide continuous assessment of information from the licensee and offsiteorganizational measurements, to initiate additional measures as indicated by actual or potential releases, toprovide consultation with offsite authorities, and to provide updates for the public through governmentauthorities.4 1P a oe 2.2 INITIATING CONDITION (IC)An event or condition that aligns with the definition of one of the four EMERGENCY CLASSIFICATIONLEVELS 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 ofan emergency classification level. An IC can be expressed as a continuous, measurable parameter (e.g., RCSleakage), an event (e.g., an earthquake) or the status of one or more fission product barriers (e.g., loss of the RCSbarrier).Appendix I of NUREG-0654 does not contain example Emergency Action Levels (EALs) for each ECL, butrather Initiating Conditions (i.e., plant conditions that indicate that a radiological emergency, or events that couldlead to a radiological emergency, has occurred). NUREG-0654 states that the Initiating Conditions forn the basisfor 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 EMERGENCYCLASSIFICATION 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 statusindications; observable events; results of calculations and analyses; entry into particular procedures; and theoccurrence of natural phenomena.2.2.2 FISSION PRODUCT BARRIER THRESHOLDA pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission productbarrier.Discussion: Fission product barrier thresholds represent threats to the defense in depth design concept thatprecludes the release of radioactive fission products to the environment. This concept relies on multiple physicalbarriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactivefission products to the environment. The primary fission product barriers are:* Fuel Clad* Reactor Coolant System (RCS)* ContainmentUpon determination that one or more fission product barrier thresholds have been exceeded, the combination ofbarrier loss and/or potential loss thresholds is compared to the fission product barrier IC/EAL criteria todetermine the appropriate ECL.5 1 P age In some accident sequences, the ICs and EALs presented in the Abnormal Radiation Levels/ RadiologicalEffluent (R) Recognition Category will be exceeded at the same time, or shortly after, the loss of one or morefission product barriers. This redundancy is intentional as the former ICs address radioactivity releases that resultin certain offsite doses from whatever cause, including events that might not be fully encompassed by fissionproduct 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 SCHEME3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS)An effective emergency classification scheme must incorporate a realistic and accurate assessment of risk, bothto plant workers and the public. There are obvious health and safety risks in underestimating the potential oractual 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 anappropriate balance between reasonably anticipated event or condition consequences, potential accidenttrajectories, and risk avoidance or minimization.There are a range of "non-emergency events" reported to the US Nuclear Regulatory Commission (NRC) staff inaccordance with the requirements of 10 CFR § 50.72. Guidance concerning these reporting requirements, andexample 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 theattributes of each ECL. The goal of this process is to answer the question, "What events or conditions should beplaced tinder each ECL?" The following sources provided information and context for the development of ECLattributes.* 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 STPEGSThe 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 Directorunderstand 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 thatinvolves:(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 theplant.(C) A consequence otherwise significant enough to warrant notification to local, State and Federalauthorities.3.1.2 ALERTAn 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 productbarrier.(B) An event or condition that significantly reduces the margin to a loss or potential loss of the fuel clador RCS fission product barrier.(C) A significant loss of control of radioactive materials resulting in an inability to control radiationlevels within the plant, or a release of radioactive materials to the environment that could result indoses 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 conditionthat 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 productbarriers within a relatively short period of time. Precursor events and conditions of this type includethose 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% ofan 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 conditionthat 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 productbarriers. Precursor events and conditions of this type include those that lead directly to core damageand loss of containment integrity.(C) A release of radioactive materials to the environment that could result in doses greater than an EPAPAG at or beyond the site boundary.(D) A HOSTILE ACTION resulting in the loss of key safety functions (reactivity control, corecooling/RPV water level or RCS heat removal) or damage to spent fuel.3.1.5 Risk-Informed InsightsEmergency preparedness is a defense-in-depth measure that is independent of the assessed risk from anyparticular accident sequence; however, the development of an effective emergency classification scheme canbenefit from a review of risk-based assessment results. To that end, the development and assignment of certainICs and EALs also considered insights from several site-specific probabilistic safety assessments (PSA -alsoknown 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 coredamage frequency. For this reason, a loss of all AC power for greater than 15 minutes, with the plantat or above Hot Shutdown, was assigned an ECL of SITE AREA EMERGENCY. Precursor events toa 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 Guide1.155, Station Blackout, may be used to determine a time-based criterion to demarcate between aSITE AREA EMERGENCY and a GENERAL EMERGENCY. The time dimension is critical to aproperly anticipatory emergency declaration since the goal is to maximize the time available for Stateand local officials to develop and implement offsite protective actions. STP is an Alternate AC plantand a Station Blackout battery copying analysis is not required. Nonetheless, a 125 VDC BatteryFour Hour Coping Analysis was conducted and provides a basis for the time-based escalation pathfrom a SITE AREA EMERGENCY to a GENERAL EMERGENCY.2. For severe core damage events, uncertainties exist in phenomena important to accident progressionsleading to containment failure. Because of these uncertainties, predicting the status of containmentintegrity may be difficult under severe accident conditions. This is why maintaining containmentintegrity alone following sequences leading to severe core damage is an insufficient basis for notescalating to a GENERAL EMERGENCY.3. PSAs indicated that leading contributors to latent fatalities were sequences involving a containmentbypass, a large Loss of Coolant Accident (LOCA) with early containment failure, a Station Blackoutlasting longer than four hours, and a reactor coolant pump seal failure. The generic EAL methodologyneeds to be sufficiently rigorous to address these sequences in a timely fashion.91 P age 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTIONLEVELSThe STPEGS methodology makes use of symptom-based, barrier-based and event-based ICs and EALs. Eachtype is discussed below.Symptom-based ICs and EALs are parameters or conditions that are measurable over some range using plantinstrumentation (e.g., core temperature, reactor coolant level, radiological effluent, etc.). When one or more ofthese parameters or conditions are off-normal, reactor operators will implement procedures to identify theprobable cause(s) and take corrective action.Fission product barrier-based ICs and EALs are the subset of symptom-based EALs that refer specifically to thelevel of challenge to the principal barriers against the release of radioactive material from the reactor core to theenvironment. These barriers are the fuel cladding, the reactor coolant system pressure boundary, and thecontainment. 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 safetysignificance. These include the failure of an automatic reactor scram/trip to shut down the reactor, naturalphenomena (e.g., an earthquake), or man-made hazards such as a toxic gas release.3.3 STPEGS DESIGN CONSIDERATIONSThe South Texas Project Electrical Generating Station (STPEGS) is composed of two units, each having anidentical 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 thecoolant, 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 heatgenerated in the core to the SGs using RCPs to circulate the water. RCS pressure is maintained by means of apressurizer attached to the hot leg of one of the loops. The RCS is designed to circulate borated demineralizedwater at temperatures, pressures and flow rates consistent with the design thermal and hydraulic performance ofthe NSSS.The Reactor Coolant Pressure Boundary Leak Detection System consists of temperature, level, humidity, andradioactivity sensors with associated instrumentation and alarms. Small leaks are detected by temperature andlevel changes of systems, increasing sump levels, and humidity and radioactivity concentration changes insidethe Containment. Large leaks are detected by changes in reactor coolant inventory, changes in flow rates inprocess lines and changes in sump level.Emergency Core Cooling System consists of three independent trains, each one capable of providing 100 percentof the required flow to the core in the unlikely event of a LOCA. Each train consists of one high-head safetyinjection pump and one low-head safety injection pump. Heat is removed from the system during recirculationby the residual heat removal heat exchanger (low-head pump only). The piping and valving associated with eachof the three subsystems are identical. In the event of a steam pipe rupture, the ECCS provides adequate shutdowncapability.10 I P a g e The Reactor Containment is a post-tensioned concrete cylinder with a steel liner plate, hemispherical top, andflat bottom. This structure provides a virtually leaktight barrier to prevent escape of fission products to theenvironment in the unlikely event of a loss of coolant accident (LOCA).3.4 ORGANIZATION AND PRESENTATION OF GENERIC INFORMATIONThe 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 11Each Recognition Category section contains a matrix showing the ICs and their associated EMERGENCYCLASSIFICATION 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 byfission product barrier and the degree of barrier challenge (i.e., potential loss or loss). This presentationmethod 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 somecases, the basis also includes relevant source information and references.3.5 IC AND EAL MODE APPLICABILITYThe STPEGS emergency classification scheme was developed recognizing that the applicability of ICs andEALs will vary with plant mode. For example, some symptom-based ICs and EALs can be assessed only duringthe power operations, startup, or hot standby/shutdown modes of operation when all fission product barriers arein place, and plant instrumentation and safety systems are fully operational. In the cold shutdown and refuelingmodes, different symptom-based ICs and EALs will come into play to reflect the opening of systems for routinemaintenance, 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 EALsfor a given Recognition Category are applicable in the indicated modes.I I I P a e MODE OF APPLICABILITY MATRIXRecognition CategorvMode R C E F H SPower Operations X X X X XStartup X X X X XHot Standby X X X X XHot Shutdown X X X X XCold Shutdown X X X XRefueling X X X XDefueled X X X XSTPEGS Operating ModesMode Description Criteria (Rx Power excludes decay heat)1 Power Operations [ Reactor Power> 5%, Keff> 0.99 T Avg> 350'F2 Startup [Reactor Power < 5%, Keff> 0.99 T Avg > 350'F[ Hot Standby Reactor Power 0% Keff< 0.99 T Avg> 350'F4 Hot Shutdown Reactor Power 0% Keff < 0.99 350'F > T Avg > 200'F5 Cold Shutdown Reactor Power 0% Keff < 0.99 T Avg < 200'F6 Refueling Reactor Power 0% Keff< 0.95 T Avg < 140°FFuel in the reactor vessel with the vessel head closure bolts less than fully tensionedor with the head removed.Defueled All fuel removed from the reactor vessel (i.e., full core offload during refuel orextended outage)12 1 P ag- e 4 STPEGS SCHEME DEVELOPMENT4.1 GENERAL DEVELOPMENT PROCESSTile STPEGS ICs and EALs were developed to be unambiguous and readily assessable because both servespecific 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 developedwith input from key stakeholders such as Operations, Training, Health Physics, and Engineering. STPEGSspecific indications, parameters and values were consistent with licensing basis documents, plant procedures,training, calculations, and drawingsUseful acronyms and abbreviations associated with the STPEGS emergency classification scheme are presentedin Appendix A, Acronyms and Abbreviations. Those specific to STPEGS were included to be consistent withsite 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 arepresented in Appendix B, Definitions.4.2 CRITICAL CHARACTERISTICSWhen crafting the scheme, STPEGS ensured that certain critical characteristics were met. These criticalcharacteristics are listed below.The ICs,. EALs, Operating Mode Applicability criteria, Notes and Basis information are consistent withindustry guidance; while the actual wording may be different from NEI 99-01 Revision 6, the classificationintent is maintained. With respect to Recognition Category F, the STPEGS scheme included a user-aid tofacilitate 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 andare 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 EALSSTPEGS incorporated instrumentation that is reliable and routinely maintained in accordance with site programsand procedures. Alarms referenced in EAL statements are those that are the most operationally significant for thedescribed 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 maynot be readily differentiated from an instrument failure. If instrumentation failures occur that have EALsassociated with them (i.e., process radiation monitors) compensatory means of implementation may be used asdescribed in plant procedures.13 1 P a g e 4.4 REFERENCES TO STPEGS AOPS AND EOPSSome of the criteria/values used in several EALs and fission product barrier thresholds were drawn fromSTPEGS AOPs and EOPs. This approach was intended to maintain good alignment between operationaldiagnoses and emergency classification assessments. STPEGS verified the appropriate administrative controlsare in place to ensure that a subsequent change to anl AOP or EOP is screened to determine if an evaluationpursuant to 10 CFR 50.54(q) is required.5 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS5.1 GENERAL CONSIDERATIONSWhen making anl emergency classification, the Emergency Director must consider all information having abearing 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 theRecognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholdsserve the same function as an EAL.NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare anemergency condition within 15 minutes after the availability of indications to plant operators that an emergencyaction level has been exceeded and to promptly declare the emergency condition as soon as possible followingidentification of the appropriate emergency classification level. The NRC staff has provided guidance onimplementing this requirement in NSIR/DPR-ISG-0 1, Interim Staff Guidance, Emergencv Planning for NuclearPower Plants.All emergency classification assessments should be based upon VALID indications, reports or conditions. AVALID indication, report, or condition, is one that has been verified through appropriate means such that there isno 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 redundantindicators, or direct observation by plant personnel. The validation of indications should be completed in amanner that supports timely emergency declaration.For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.). the EmergencyDirector should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiologicalrelease is detected and the release start time is unknown, it should be assumed that the release duration specifiedin 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 notwarrant an emergency declaration provided that 1) the activity proceeds as planned and 2) the plant remainswithin 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 operating14 1P ag e license provided that the activity proceeds and concludes as expected. Events or conditions of this type may besubject 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 aspecific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak ratecalculation, etc.); the EAL and/or the associated basis discussion will identify the necessary analysis. In thesecases, the 15-minute declaration period starts with the availability of the analysis results that show the thresholdto be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees toestablish 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 maywarrant emergency classification, a provision for classification based on operator/management experience andjudgment is still necessary. This scheme provides the Emergency Director with the ability to classify events andconditions 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 consequencesof the event or condition reasonably meet or exceed a particular ECL definition. A similar provision isincorporated into the Fission Product Barrier Tables; judgment may be used to determine the status of a fissionproduct barrier.5.2 CLASSIFICATION METHODOLOGYTo make an emergency classification, the user will compare an event or condition (i.e., the relevant plantindications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of anEAL(s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met orexceeded, 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 timeduration runs concurrently with the emergency classification process "clock." For a full discussion of this timingrequirement, refer to NSIR/DPR-ISG-01.5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONSWhen multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. Thehighest 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 differentunits, 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 RegulatoryIssue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During QuicklyChanging Events.15 1 P age 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATIONThe mode in effect at the time that all event or condition occurred, and prior to any plant or operator response, isthe mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in amode change before the emergency is declared, the emergency classification level is still based on the mode thatexisted at the time that the event or condition was initiated (and not when it was declared). Once a different modeis reached, any new event or condition, not related to the original event or condition, requiring emergencyclassification should be evaluated against the ICs and EALs applicable to the operating mode at the time of thenew event or condition.For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the ColdShutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plantresponse. In particular, the fission product barrier EALs are applicable only to events that initiate in the HotShutdown mode or higher.5.5 CLASSIFICATION OF IMMINENT CONDITIONSAlthough EALs provide specific thresholds, the Emergency Director must remain alert to events or conditionsthat could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECLis IMMINENT). If, in thejudgment of the Emergency Director, meeting an EAL is IMMINENT, the emergencyclassification should be made as if the EAL has been met. While applicable to all EMERGENCYCLASSIFICATION LEVELS, this approach is particularly important at the higher EMERGENCYCLASSIFICATION LEVELS since it provides additional time for implementation of protective measures.5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING ANDDOWNGRADINGAn 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, thenew 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 LongerExistsUNUSUAL EVENT Terminate the emergency in accordancewith plant procedures.ALERT Downgrade or terminate the emergency inaccordance with plant procedures.SITE AREA EMERGENCY with no long- Downgrade or terminate the emergency interm plant damage accordance with plant procedures.SITE AREA EMERGENCY with long- Terminate the emergency and enterterm plant damage recovery in accordance with plantprocedures.GENERAL EMERGENCY Terminate the emergency and enterrecovery in accordance with plantprocedures.16 1 P ag e As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS2007-02.5.7 CLASSIFICATION OF SHORT-LIVED EVENTSAs discussed in Section 3.2, event-based ICs and EALs define a variety of specific occurrences that havepotential or actual safety significance. By their nature, some of these events may be short-lived and, thus, overbefore the emergency classification assessment can be completed. If an event occurs that meets or exceeds anEAL, 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 reactorfollowed by a successful manual scram/trip or an earthquake.5.8 CLASSIFICATION OF TRANSIENT CONDITIONSMany of the ICs and/or EALs contained in this document employ time-based criteria. These criteria will requirethat 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 anEAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance shouldbe applied to the classification of these conditions.EAL momentarily met during expected plant response -In instances where an EAL is briefly met during anexpected (normal) plant response, an emergency declaration is not warranted provided that associated systemsand 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 takesprompt manual action to address a condition, and the action is successful in correcting the condition prior to theemergency declaration, then the applicable EAL is not considered met and the associated emergency declarationis not required. For illustrative purposes, consider the following example.An ATWS occurs and the auxiliary feedwater system fails to automatically start. Steam generator levels rapidlylower and the plant enters an inadequate RCS heat removal condition (a potential loss of both the fuel clad andRCS barriers). If an operator manually starts the auxiliary feedwater system in accordance with an EOP step andclears the inadequate RCS heat removal condition prior to an emergency declaration, then the classificationshould 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 obviatethe need to classify the event; emergency classification assessments must be deliberate and timely, with no unduedelays. The provision discussed above addresses only those rapidly evolving situations where an operator is ableto take a successful corrective action prior to the Emergency Director completing the review and steps necessaryto make the emergency declaration. This provision is included to ensure that any public protective actionsresulting from the emergency classification are truly warranted by the plant conditions.17 ..aLe 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT ORCONDITIONIn some cases, an EAL may be met but the emergency classification was not made at the time of the event orcondition. This situation can occur when personnel discover that an event or condition existed which met anEAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. Thismay be due to the event or condition not being recognized at the time or an error that was made in the emergencyclassification process.In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 isapplicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 within onehour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State andlocal agencies in accordance with the agreed upon arrangements.5.10 RETRACTION OF AN EMERGENCY DECLARATIONGuidance 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/EALSTable R-1: Recognition Category "R" Initiating Condition MatrixUNUSUAL EVENTRUl Release of gaseousor liquid radioactivitygreater than 2 times theODCM limits for 60minutes or longer.Op. Modes: AllRU2 UNPLANNEDloss of water level aboveirradiated fuel.Op. Modes: AllALERTSITE AREAEMERGENCYGENERALEMERGENCYRA1 Release of gaseousor liquid radioactivityresulting in offsite dosegreater than 10 mrernTEDE or 50 mremTHYROID CDE.Op. Modes: AllRA2 Significantlowering of water levelabove, or damage to,irradiated fuel.Op. Modes: AllRA3 Radiation levelsthat impede access toequipment necessary fornonnal plant operations,cooldown or shutdown.Op. Modes: AllRS1 Release of gaseousradioactivity resulting inoffsite dose greater than100 mrem TEDE or 500mrern THYROID CDE.Op. Modes: AllRS2 Spent fuel poollevel at 40'-4" or lower.Op. Modes: AllRG1 Release of gaseousradioactivity resulting inoffsite dose greater than1,000 mrem TEDE or5,000 mrem THYROIDCDE.Op. Modes: AllRG2 Spent fuel poollevel cannot be restoredto at least 40'-4" for 60minutes or longer.Op. Modes: All191 Page RU1ECL: UNUSUAL EVENTInitiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for60 minutes or longer.Operating Mode Applicability: ALLEmergency Action Levels: (1 or 2 or 3)Notes:* The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that 60 minuteshas 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 hasexceeded 60 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the releasepath, 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 MonitorsRelease Point Monitor GE SAE ALERT UEUnit 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/secMain Steam RT-8046 thni 4.00 E+02 ptCi/cm3 4.00 E+01 pCi/cm3 4.00 E+00 ptCi/cm3 5.00 E-02 pCi/cm3Lines 8049(2) Reading on gaseous effluent radiation monitor RT-801 OB greater than 2 times the alarm setpointestablished 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 2times 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 radiologicalrelease that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). Itincludes any gaseous or liquid radiological release, monitored or un-monitored, including those for which aradioactivity 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 andmonitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environmentis 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 cannotbe readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant conditionand 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. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, thenthe 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 30minutes does not meet the EAL.EAL #1- This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous orliquid effluent pathways.EAL #2- This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2times the limit established by a radioactivity discharge permit. This EAL will typically be associated withplanned 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 orenvironmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into stormdrains, heat exchanger leakage in river water systems, etc.).Escalation of the emergency classification level would be via IC RAI.RUI: EAL-1 Selection BasisThe Unit Vent and Main Steam Line monitor readings were included in this EAL because they giveinstantaneous indications of a monitored gaseous release exceeding twice the ODCM limits. Normal gaseouseffluents are due to planned RCB purges and monitored by the Unit Vent. The Main Steam Line monitorreadings were included because they correspond to a concentration that would result in a release rate of twice theODCM limits if there were a release via the Power Operated Relief Valves (PORVs) or Safety Relief Valves. Arelease from the PORVs or Safety Relief Valves is not a normal effluent pathway but ineets the intent of theEAL.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 BasisFor EAL-2, there are two effluent radiation monitors, RT-8038 (liquid) and RT-800 OB (gaseous), however onlyRT-8010B was included. The alarm setpoint for the gaseous effluent radiation monitor RT-8010B is set at theODCM limits. An indication of two times the alarm setpoint (two times the ODCM limit) would allow operatorstime to secure the release prior to meeting this EAL. The liquid effluent radiation monitor RT-8038 was notincluded in EAL-2 because the activity in liquid discharges is normally the several orders of magnitude lowerthan the ODCM limits. In order to alert personnel to significant changes in the liquid effluent activity, the alarmsetpoint for RT-8038 is normally set several orders of magnitude below the ODCM limits. Setting the alarmsetpoint 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 EmergencyAction Levels2. Offsite Dose Calculation Manual (ODCM), Rev. 17, Part B3.0 to B4.93. 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 RU2ECL: UNUSUAL EVENTInitiating Condition: UNPLANNED loss of water level above irradiated fuel.Operating Mode Applicability: ALLEmergency Action Level:(1) a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of thefollowing:* Visual ObservationOR* 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 LevelHI/LO) is in alarmANDb. UNPLANNED rise in area radiation levels on ANY of the following radiation monitors.* RE-8055 (68' RCB) -Mode 5 or 6 onlyOR" RE-8099 (68' RCB) -Mode 5 or 6 onlyOR" 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 tocontrol 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. Othersources of level indications may include reports from plant personnel (e.g., from a refueling crew) or videocamera observations. A significant drop in the water level may also cause a rise in the radiation levels of adjacentareas that can be detected by monitors in those locations.The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitorreading may rise due to planned evolutions such as lifting of the reactor vessel head or movement of a fuelassembly. Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNEDloss 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 RecognitionCategory C during the Cold Shutdown and Refueling modes.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via IC RA2.RU2: EAL-1 Selection BasisHi/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 thenormal water level of 66'-6" (UFSAR 9.1.2.1), the "SFP WATER LEVEL H I/LO" lampbox 22M02 window F-5annunciator alarm is received in the Control Room (0POP09-AN-22M2, Annunciator Lampbox 22M02Response Instructions).Although the ICSA has a Hi/LO level sensor, there is not an annunciator in the Control Room similar to the onefor the Spent Fuel Pool. There is however, a "SFP TROUBLE" lampbox 22M02 window F-6 annunciator in thecontrol room. One of the inputs to this alarm is FC-LSHL-1420, the ICSA HI/LO level sensor. Since no fuel islocated 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 bioshieldwall close to the refueling cavity. Area radiation monitor RE-8090 is located in the Fuel Handling Building on68' 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 thetop of any array, the dose rate would be expected not to exceed 2.5 mR/hr, above background. This assumes 42hours 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 fromseveral hundred R/hr to over a thousand R/hr on and around the 68' elevation deck (table C-5 NUREGCR/0649).REFERENCES:I. OPOP09-AN-22M2, Rev. 25, Annunciator Lampbox 22M02 Response Instructions F-5 and F-6Window (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 Monitors24 1 P a-, e RA1ECL: ALERTInitiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10mrem TEDE or 50 mrem THYROID CDE.Operating Mode Applicability: ALLEmergency Action Levels: (I or 2 or 3 or 4)Notes:* The Emergency Director should declare the ALERT promptly upon determining that the applicable time hasbeen exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration hasexceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the releasepath, 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 MonitorsRelease Point Monitor GE SAE ALERT UEUnit 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/secMain Steam RT-8046 thru 4.00 E+02 pCi/cm3 4.00 E+01 gCi/cm3 4.00 E+00 pLCilcrn3 5.00 E-02 pCi/cm3Lines 8049(2) Dose assessment using actual meteorology indicates doses greater than 10 mremn TEDE or 50 mremTHYROID CDE at or beyond the SITE BOUNDARY.(3) Analysis of a liquid effluent sample indicates a concentration or release rate that would result in dosesgreater than 10 mrem TEDE or 50 mrem THYROID CDE at or beyond the SITE BOUNDARY for onehour 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 ofinhalation.25 1P a e Basis:This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite dosesgreater than or equal to 1% of the EPA PROTECTIVE ACTION GUIDES (PAGs). It includes both monitoredand un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation ofthe 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 cannotbe readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant conditionand 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 wasestablished 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. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, thenthe 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 BasisThe Unit Vent and Main Steam Line monitor readings were included in this EAL because they giveinstantaneous indications of a monitored gaseous release meeting the EAL threshold values of 10 mremTEDE or 50 mrem CDE THYROID at the SITE BOUNDARY. Gaseous releases from the plant aremonitored by the Unit Vent. The Main Steam Line monitor readings correspond to a concentration thatwould result in a release rate meeting the EAL threshold values if there were a release via the Power OperatedRelief 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 calculatedvalues to ensure they are readily assessable.RAI: EAL-2, 3, 4 Selection BasisN/AREFERENCES:1. Calculation No: STPNOC013-CALC-002 Rev. 2., Radiological Release Thresholds for EmergencyAction Levels2. 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 RA2ECL: ALERTInitiating Condition: Significant lowering of water level above, or damage to, irradiated fuel.Operating Mode Applicability: ALLEmergency 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 thefollowing FHB radiation monitor readings:* FHB Exhaust, RT-8035 or RT-8036 greater than 1.00 E- I.tCi/cm3OR* ARM (68' FHB), RE-8090 greater than 1,500 mR/hrORb. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of thefollowing RCB radiation monitor readings (Mode 5 or 6 only).* ARMs (68' RCB), RE-8055 or RE-8099 greater than 850 mR/hr.NOTEEAL-3 is not applicable until the enhanced SFPlevel 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 asignificant lowering of water level within the spent fitel pool or Inside Containment Storage Area (UCSA).Theseevents present radiological safety challenges to plant personnel and are precursors to a release of radioactivity tothe environment. As such, they represent an actual or potential substantial degradation of the level of safety ofthe plant.This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask issealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified inaccordance 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 REFUELINGPATHWAY, is of sufficient magnitude to have resulted in Uncovery of irradiated fuel. Indications of irradiatedfuel 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 mayalso be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality ofavailable indications, reports and observations. While an area radiation monitor could detect a rise in a dose ratedue to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be areliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should beconsidered 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 RecognitionCategory 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 ontoan assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reportsor 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 preventsignificant dose consequences from direct gamma radiation to personnel performing operations in the vicinity ofthe spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventoly and thus it is alsoa 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 isbased on Calculation STPNOCOI3-CALC-005 Rev.2. The threshold value of 1500 mR/hr for area radiationmonitor RE-8090 was truncated less than 4% from the calculated value to ensure the threshold was readilyassessable. Threshold values for FHB Exhaust Monitors RT-8035 and RT-8036 were also included because theyare accident monitors that are sensitive to noble gases which are expected to be present if irradiated fuel isdamaged. The calculated monitor reading for RT-8035 and RT-8036 is 3.8 ýtCi/cm3 and the high range of themonitors is 0.3 [tCi/cm3.The threshold value of 0.1 pCi/cm3 is approximately 6 orders of magnitude abovebackground and indicative of damaged irradiated fuel. It was selected because it is readily assessable and withinthe calibrated range of the monitors.The calculated airborne source term and radiation monitor response for a fuel handling accident in the RCB isbased 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 Regardto 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 substantialradiation 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 12-02indicates 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 2criteria as 49'- 10".REFERENCES:1. Calculation No.: STPNOCOI3-CALC-005 Rev.2, Fuel Handling Accident Monitor Response forEmergency 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 FuelPool Level histrumentation to Meet NRC Order EA- 12-051, Rev. 0, NOC -AE- 1300295929 1P a ge RA3ECL: ALERTInitiating Condition: Radiation levels that impede access to equipment necessary for normal plant operations,cooldown or shutdown.Operating Mode Applicability: ALLEmergency Action Levels: (1 or 2)Note: If the equipment in the listed room or area was already inoperable or out-of-service before the eventoccurred, 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 areaslisted in Table H3/R2:__.___ ;TABLE H3/R2: Plant'Areas'l.Rquiring AccessRCB RHR Heat Exchanger Rooms00 MAB 51 ft Room 335EAB Roof, MCC 1G8, 4.16KV Switchgear RoomsLU00 "n EAB 4. 16KV Switchgear Rooms30 1 P a g e Basis:This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impedepersonnel from performing actions necessary to maintain normal plant operation, or to perform a normal plantcooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level ofsafety of the plant. The Emergency Director should consider the cause of the higher radiation levels anddetermine 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, procedurallyrequired during the plant operating mode in effect at the time of the elevated radiation levels. The emergencyclassification 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 personnelinto the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protectiveequipment, 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., entryis not required during the operating mode in effect at the time of the elevated radiation levels). Forexample, the plant is in Mode I when the radiation rise occurs, and the procedures used for normaloperation, 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 whichaddress 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 actuallyprevent 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 forexpected occupancy times. The rooms listed in EAL-1 require continuous occupancy to maintain normal plantoperation, 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, orshutdown. Assuming all plant equipment is operating as designed, Normal operations and safe shutdownequipment operation is capable from the Main Control Room (MCR). The plant is able to transition into a hotshutdown from the MCR, therefore H3/R2 is a list of plant rooms or areas with entry-related mode applicabilitythat 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, andlimit dilution pathways) and subsequent entry into cold shutdown (disable operation of ECCS equipment). Afterachieving 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 192. OPOP03-ZG-0008, Rev. 56, Power Operations3. OPOP03-ZG-0006, Rev. 54, Plant Shutdown from 100% to Hot Standby4. OPOP03-ZG-0007, Rev. 7 1, Plant Cooldown32 1 P a ge RS1ECL: SITE AREA EMERGENCYInitiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 rnrern TEDE or500 mrem THYROID CDE.Operating Mode Applicability: ALLEmergency Action Levels: (I or 2 or 3)Notes:The Emergency Director should declare the SITE AREA EMERGENCY promptly upon determining that theapplicable 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 hasexceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the releasepath, 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 classificationassessments 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 MonitorsRelease Point Monitor GE SAE ALERT UEUnit Vent RT-8010B 1.50 E+08 1.50 E+07 pCi/sec 1.50 E+06 1.40 E+05 ýtCi/secMain Steam RT-8046 thru 4.00 E+02 pCi/cm3 4.00 E+01 pCi/cm3 4.00 E+00 ptCi/cm3 5.00 E-02 pCi/cm3Lines 8049(2) Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mremTHYROID 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 ofinhalation.33' I P a ge Basis:This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than orequal 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 theprotection of the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that carmotbe readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant conditionand 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 wasestablished 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. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, thenthe 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 giveinstantaneous indications of a monitored gaseous release meeting the EAL threshold values of 100 mremTEDE or 500 mremn CDE THYROID at the SITE BOUNDARY. Gaseous releases from the plant aremonitored by the Unit Vent. The Main Steam Line monitor readings correspond to a concentration thatwould result in a release rate meeting the EAL threshold values if there were a release via the Power OperatedRelief Valves (PORVs) or Safety Relief Valves.The Unit Vent and Main Steam Line release values are based on Calculation No. STPNOC013-CALC-002Rev.2. The adjusted values used in this EAL were conservatively truncated by less than 1% of the calculatedvalues to ensure they are readily assessable.RS1: EAL-2. EAL-3 Selection Basis:N/AREFERENCES:1. Calculation No: STPNOCO I 3-CALC-002 Rev.2, Radiological Release Thresholds for EmergencyAction Levels2. UFSAR Section, Rev. 14, Section 11.5.2.3.3 and 11.5.2.5.3 (monitor descriptions)34 1 P a g e RS2ECL: SITE AREA EMERGENCYInitiating Condition: Spent fuel pool level at 40'-4" or lower.Operating Mode Applicability: ALLEmergency Action Level:NOTEEAL-I is not applicable until the enhanced SFPlevel 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 toIMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of thepublic 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 wasmet; 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 Regardto 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 actionsto 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 Level3 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 ModifyLicenses with Regard to Reliable Spent Fuel Pool Instrumentation", August 20121. South Texas Project (STP) Overall Integrated Plan for Implementation of Unit I & Unit 2 Spent FuelPool Level Instrumentation to Meet NRC Order EA-12-051, Rev. 0, NOC -AE-1300295935 I P a ,, e RG1ECL: GENERAL EMERGENCYInitiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDEor 5,000 mrern THYROID CDE.Operating Mode Applicability: ALLEmergency Action Levels: (1 or 2 or 3)Notes:* The Emergency Director should declare the GENERAL EMERGENCY promptly upon determining that theapplicable 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 hasexceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the releasepath, 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 classificationassessments 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 MonitorsRelease Point Monitor GE SAE ALERT LIEUnit 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/secMain Steam RT-8046 thni 4.00 E+02 iLCi/cm3 4.00 E+01 piCi/cm3 4.00 E+00 lpCi/cm 3 5.00 E-02 pCi/cm3Lines 8049(2) Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000mrem 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 ofinhalation.36 1P ag-e Basis:This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than orequal to the EPA PROTECTIVE ACTION GUIDES (PAGs). It includes both monitored and un-monitoredreleases. 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 cannotbe readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant conditionand 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 inconsideration 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. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, thenthe 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 giveinstantaneous indications of a monitored gaseous release meeting the EAL threshold values of 1000 mrem TEDEor 5000 mrem CDE THYROID at the SITE BOUNDARY. Gaseous releases from the plant are monitored by theUnit Vent. The Main Steam Line monitor readings correspond to a concentration that would result in a releaserate meeting the EAL threshold values if the release was via the Power Operated Relief Valves (PORVs) orSafety Relief Valves.The Unit Vent and Main Steam Line release values are based on Calculation No. STPNOCOI3-CALC-002Rev.2. The adjusted values used in this EAL were conservatively truncated by less than 1% of the calculatedvalues to ensure they are readily assessable.RGI: EAL-2. EAL-3 Selection Basis:N/AREFERENCES:1. Calculation No: STPNOCO13-CALC-002 Rev.2, Radiological Release Thresholds for EmergencyAction Levels,2. STP UFSAR, Rev. 14, Section 11.5.2.3.3 and 11.5.2.5.3 (monitor descriptions)371 P a ge RG2ECL: GENERAL EMERGENCYInitiating Condition: Spent fuel pool level cannot be restored to at least 40'-4" for 60 minutes or longer.Operating Mode Applicability: ALLEmergency Action Level:Note: The Emergency Director should declare the GENERAL EMERGENCY promptly upon determining that60 minutes has been exceeded, or will likely be exceeded.NOTEEAL-1 is not applicable until the enhanced SFPlevel 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 aprolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to theenvironment.It is recognized that this IC would likely not be met until well after another GENERAL EMERGENCY IC wasmet; 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 withRegard 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 actionsto 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 3criteria 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 ModifyLicenses with Regard to Reliable Spent Fuel Pool Instrumentation", August 20124. South Texas Project (STP) Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent FuelPool Level Instrumentation to Meet NRC Order EA-12-051, Rev. 0, NOC -AE-130029593911-1ag e 7 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONICS/EALSTable C-i: Recognition Category "C" Initiating Condition MatrixUNUSUAL EVENTCU1 UNPLANNEDloss of RCS inventoryfor 15 minutes orlonger.Op. Modes: 5,6CU2 Loss of ALL butone AC power source toemergency buses for 15minutes or longer.Op. Modes: 5,6DefueledCU3 UNPLANNEDrise in RCS temperature.Op. AModes: 5,6CU4 Loss of Vital DCpower for 15 minutes orlonger.Op. Modes: 5,6ALERTSITE AREAEMERGENCYGENERALEMERGENCYCA1 Loss of RCSinventory.Op. Modes. 5,6CS1 Loss of RCSinventory affecting coredecay heat removalcapability.Op. Modes: 5,6CG1 Loss of RCSinventory affecting fuelclad integrity withcontainment challenged.Op. Modes: 5,6CA2 Loss of ALLoffsite and ALL onsiteAC power to emergencybuses for 15 minutes orlonger.Op. Modes: 5,6,DefueledCA3 Inability tomaintain the plant incold shutdown.Op. Modes: 5,6CU5 Loss of ALLonsite or offsitecommunicationscapabilities.Op. Modes: 5,6,DefuieledCA6 Hazardous eventaffecting a SAFETYSYSTEM needed forthe current operatingmode.Op. Modes: 5,640 Pa e culECL: UNUSUAL EVENTInitiating Condition: UNPLANNED loss of RCS inventory for 15 minutes or longer.Operating Mode Applicability: 5, 6Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that 15minutes has been exceeded, or will likely be exceeded.(1) UNPLANNED loss of reactor coolant results in RCS level below the procedurally required limit for 15minutes or longer.(2) a. RCS level cannot be monitored.ANDb. 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 SumpBasis:This IC addresses the inability to restore and maintain water level to a required minimum level (or the lowerlimit 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 UNPLANNEDevent that results in water level decreasing below a procedurally required limit warrants the declaration of anUNUSUAL 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 arefueling outage as different plant configurations and system lineups are implemented. This EAL is met if theminimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. Theminimum 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 theexpected 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 theyare indicative of leakage from the RCSContinued loss of RCS inventory may result in escalation to the ALERT EMERGENCY CLASSIFICATIONLEVEL 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., headremoval, mid-loop operations) as described in OPOP03-ZG-0009, Mid-Loop Operation. The 15 minute timeframe allows for prompt operator actions to restore RCS level in the event of an UNPLANNED lowering of RCSlevel 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 secondcondition 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, ExcessiveRCS Leakage. Since other system leaks could rise levels in various tanks and sumps, the list has been limited tothe tanks and sumps that would have the highest probability of indicating RCS leakage inside the ReactorContainment 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 Leakage2. OPOP03-ZG-0007, Rev. 71, Plant Cooldown3. OPOP03-ZG-0009, Rev. 59, Mid-Loop Operation42 1 Page CU2ECL: UNUSUAL EVENTInitiating Condition: Loss of ALL but one AC power source to emergency buses for 15 minutes or longer.Operating Mode Applicability: 5, 6, DefueledEmergency Action Level:Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that 15minutes 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 15minutes or longer.ANDb. ANY additional single power source failure will result in loss of ALL AC power to SAFETYSYSTEMS.Basis:This IC describes a significant degradation of offsite and onsite AC power sources such that any additional singlefailure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC powersource 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 becauseof the additional time available to restore another power source to service. Additional time is available due to thereduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when inthese 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 toan 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 onsitediesel generator).* A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with asingle 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 busesbeing 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 accordancewith 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 anyadditional single failure would results in a loss of all AC power. This condition is an UNUSUAL EVENTduring modes 5, 6 and Deftieled because of the additional time available to restore power due to the reduced coredecay heat load, and the lower temperatures and pressures in various plant systems. In modes 1-4, this conditionis 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 Bus2. OPOP04-AE-0004, Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus3. OPSP03-EA-0002, Rev. 32, ESF Power Availability4. Drawing OOOOOEOAAAA, Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. 1 & 244 1 P a g e CU3ECL: UNUSUAL EVENTInitiating Condition: UNPLANNED rise in RCS temperature.Operating Mode Applicability: 5, 6Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that 15minutes 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 shutdowntemperature limit, or the inability to determine RCS temperature and level, represents a potential degradation ofthe level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not establishedduring this event, the Emergency Director should also refer to IC CA3.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limitwhen 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 thatwhich can currently be removed, such that reactor coolant temperature cannot be maintained below the coldshutdown temperature limit specified in Technical Specifications. During this condition, there is no immediatethreat 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 temperaturedepending on the time after shutdown.EAL #2- reflects a condition where there has been a significant loss of instrumentation capability necessary tomonitor RCS conditions and operators would be unable to monitor key parameters necessary to assure coredecay heat removal. During this condition, there is no immediate threat of fuel damage because the core decayheat 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 toALERT 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 theinability to control RCS temperature. Mode 4 (Hot Shutdown) would be entered when Tavg exceeds 200 'F(Reference 1).CU3: EAL-2 Selection Basis:N/AREFERENCES:1. Technical Specifications Table 1.2 (Mode, Temperature, Power, ketf Table)46 1 P a e CU4ECL: UNUSUAL EVENTInitiating Condition: Loss of Vital DC power for 15 minutes or longer.Operating Mode Applicability: 5, 6Emergency Action Level:Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that 15minutes 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 operableSAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decayheat load has been significantly reduced, and coolant system temperatures and pressures are lower; theseconditions extend the time available to restore a vital DC bus to service. Thus, this condition is considered to bea 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, oroperable, 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 DCpower affecting Train B would require the declaration of an UNUSUAL EVENT. A loss of Vital DC power toTrain 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 CAIor 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 tobe 105.5 volts. At 105.5 volts or less, OPOP05-EO-EC0O, Loss of All AC Power, directs the operators to openthe battery output breakers.REFERENCES:1. Calculation 13-DJ-006, Rev. 0, 125 VDC Battery Four Hour Coping Analysis2. OPOP05-EO-ECOO, Rev. 23, Loss of All AC Power47 1 P a g e CU5ECL: UNUSUAL EVENTInitiating Condition: Loss of ALL onsite or offsite communications capabilities.Operating Mode Applicability: 5, 6, DeflieledEmergency 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 TableC4.(3) Loss of ALL of the following NRC communication methods in Table C4.Table C4: Communications MethodsEAL-1 EAL-2 EAL-3ONSITE ORO NRCPlant PA system XPlant Radios XPlant telephone system X X XSatellite phones X XDirect line from Control Rooms to Bay City X XMicrowave Lines to Houston X XSecurity radio to Matagorda County XDedicated Ring-down lines XENS line XBasis:This IC addresses a significant loss of on-site or offsite communications capabilities. While not a directchallenge 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 multipleradio 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 emergencydeclaration. The OROs referred to here are Matagorda County Sheriff's Office. and Texas Department of PublicSafety 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 emergencydeclaration.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 ofinformation. Links were obtained from procedures OPGP05-ZV-001 L, Emergency Communications.REFERENCES:1. OPGP05-ZV-001 1, Rev. 8, Emergency Communications491 P a g e CA1ECL: ALERTInitiating Condition: Loss of RCS inventory.Operating Mode Applicability: 5, 6Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the ALERT promptly upon determining that 15 minutes has beenexceeded, 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 longerANDb. UNPLANNED rise in ANY of the following sump or tank levels in Table C2 due to a loss of reactorvessel/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 SumpBasis:This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., aprecursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in thelevel of plant safety.EAL #1- A lowering of water level below elevation 32'- 9" indicates that operator actions have not beensuccessful in restoring and maintaining reactor vessel/ water level. The heat-up rate of the coolant will rise as theavailable water inventory is reduced. A continuing reduction in water level will lead to core uncovery. Althoughrelated, EAL #1 is concerned with the loss of RCS inventory and.not the potential concurrent effects on systemsneeded for decay heat removal (e.g., loss of a Residual Heat Removal suction point). Arise in RCS temperaturecaused 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 powerfailures, or water level dropping below the range of available instrumentation. If water level cannot bemonitored, operators may determine that an inventory loss is occurring by observing changes in sump and/ortank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to50 P a e ensure they are indicative of leakage from the reactor vessel/RCS. The 15-minute duration for the loss of levelindication was chosen because it is half of the EAL duration specified in IC CSIIf the reactor vessel/RCS inventory level continues to lower, then escalation to SITE AREA EMERGENCYwould 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 (+ 6inches above hot leg centerline). If RCS inventory is reduced below this level, normal decay heat removalsystems may not be available for core cooling. This threshold is not applicable to reduced inventory vacuum fillsince 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. Sinceother system leaks could raise levels in various tanks and sumps, the list was limited to the tanks and sumps thatwould 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 Leakage2. OPOP02-RH-00011, Rev. 63, Residual Heat Removal System Operation51 1Page CA2ECL: ALERTInitiating Condition: Loss of ALL offsite and ALL onsite AC power to emergency buses for 15 minutes orlonger.Operating Mode Applicability: 5, 6, DefuieledEmergency Action Level:Note: The Emergency Director should declare the ALERT promptly upon determining that 15 minutes has beenexceeded, or will likely be exceeded.(1) Loss of ALL offsite AND ALL onsite AC Power to ALL three 4160V AC ESF Busses for 15 minutesor longer.Basis:This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMSrequiring electric power including those necessary for emergency core cooling, containment heatremoval/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 AREAEMERGENCY because of the additional time available to restore an emergency bus to service. Additional timeis available due to the reduced core decay heat load, and the lower temperatures and pressures in various plantsystems. Thus, when in these modes, this condition represents an actual or potential substantial degradation ofthe 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/AREFERENCES:1. OPOP04-AE-0001, Rev. 44, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus2. OPOP04-AE-0004, Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus3. OPSP03-EA-0002, Rev. 32, ESF Power Availability4. Drawing OOOOOEOAAAA, Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. I & 252 111 a ge CA3ECL: ALERTInitiating Condition: Inability to maintain the plant in cold shutdown.Operating Mode Applicability: 5, 6Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the ALERT promptly upon determining that the applicable timehas been exceeded, or will likely be exceeded.(1) UNPLANNED rise in RCS temperature to greater than 200 'F (Tavg) for greater than the durationspecified in Table C3.Table C3: RCS Heat-up Duration ThresholdsRCS Status Containment Closure Status Heat-up DurationIntact (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 isbeing reduced, the EAL is not applicable.(2) UNPLANNED RCS pressure rise greater than 10 psig. (This EAL does not apply during water-solidplant conditions.)Basis: This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to theRCS in excess of that which can currently be removed. Either condition represents an actual or potentialsubstantial degradation of the level of safety of the plant.A momentary UNPLANNED excursion above the Teclmical Specification cold shutdown temperature limitwhen 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 whenCONTAINMENT 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 thetemperature rise.The RCS Heat-up Duration Thresholds table also addresses an rise in RCS temperature with the RCS intact. Thestatus of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a highpressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to addressthe 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 andCONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently tothe 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 generatorU-tube draining, RCS level differences when operating at a mid-loop condition, decay heat removal systemdesign, and level instrumentation problems can lead to conditions where decay heat removal is lost and coreuncover can occur. NRC analyses show that there are sequences that can cause core uncovery in 15 to 20minutes, and severe core damage within an hour after decay heat removal is lost. The allowed time frames areconsistent with the guidance provided by Generic Letter 88-17 and believed to be conservative given that a lowpressure 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 determinedusing installed instrumentation, but not less than 10 psig.REFERENCES:I. Technical Specifications Table 1.2 (Mode, Temperature, Power., keff Table)5411Page CA6ECL: ALERTInitiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operatingmode.Operating Mode Applicability: 5, 6Emergency 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 retainingdike along the North Wall* Other events with similar hazard characteristics as determined bythe Shift ManagerANDb. EITHER of the following:1. Event damage has caused indications of degraded performance in at least one train of aSAFETY SYSTEM needed for the current operating mode.OR2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structureneeded for the current operating mode.Basis:This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containingSAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces themargin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potentialsubstantial 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 itwill be readily available. The indications of degraded performance should be significant enough to cause concernregarding 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 readilyapparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will55 1r a -e make this determination based on the totality of available event and damage report information. This is intendedto 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 theMain 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 Protection56 P1 a g e CS1ECL: SITE AREA EMERGENCYInitiating Condition: Loss of RCS inventory affecting core decay heat removal capability.Operating Mode Applicability: 5, 6Emergency Action Levels: (1 or 2 or 3)Note: The Emergency Director should declare the SITE AREA EMERGENCY promptly upon determining that30 minutes has been exceeded, or will likely be exceeded.(1) a. CONTAINMENT CLOSURE not established.ANDb. RCS level less than 33% of plenum.(2) a. CONTAINMENT CLOSURE established.ANDb. RCS level less than 0% of plenum(3) a. RCS level cannot be monitored for 30 minutes or longer.ANDb. Core uncovery is indicated by ANY of the following:* Reactor Containment Building, 68'-0" Area Radiation Monitors RE-8055 or RE-8099 readinggreater 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 sufficientmagnitude 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 Sump57 1 P a g e Basis:This IC addresses a significant and prolonged loss of reactor vessel/RCS inventory control and makeupcapability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, aloss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures ofplant 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 reactorcoolant 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 CLOSUREfollowing a loss of heat removal or RCS inventory control functions. The difference in the specified RCS levelsof EALs L.b and 2.b reflect the fact that with CONTAINMENT CLOSURE established, there is a lowerprobability 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 abilityto monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions todetermine if core uncovery has actually occurTed (i.e., to account for various accident progression andinstrumentation 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 leveldropping below the range of available instrumentation. If water level cannot be monitored, operators maydetermine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tanklevel changes must be evaluated against other potential sources of water flow to ensure they are indicative ofleakage 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 atCommercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actionsto 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 theRCS loop penetration at the reactor vessel. Six inches (6") below the bottom inside diameter of the RCS hot legnozzle (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 6inches above (Sensor 6) and 4.9 inches below (Sensor7) the prescribed elevation of 30'-6.5". When water level isat the desired elevation of 30'-6.5", Sensor 6 will be dry and Sensor 7 will be wet. This condition corresponds toa reading of 33% of plenum per 0POP02- 11-0002, RVWL Monitoring System, Addendum 1, RVWL SensorElevations.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 RCSlevel 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 MonitoringSystem thennocouple to TAF is Sensor 8 at elevation 29'-2.7". Use of RVWL to approximate TAF; with theinherent gap of 12 inches between indicated level and actual level, is acceptable for the purposes of signaling thatthe 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 arelocated on the 68'-0" elevation of the reactor containment building. Their locations are identified on drawing9C129A81105. Their range (0.1 mR/hr to 10,000 mR/hr) is identified in Table 12.3.4-1 of Section 12 of theUFSAR. 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 coreuncovery.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 fueland 6 R/hr at one foot above the top of active fuel. The high range of these monitors is 10 R/hr. The value of9,000 mR/hr was selected to ensure that the threshold is readily assessable and within the calibrated range of themonitor. The threshold value of 9,000 mnR/hr corresponds to approximately 8 inches above the top of the activefuel with the reactor head on; which provides an additional indication that RCS levels are near the point of fueluncovery. These monitor readings in conjunction with the other threshold values allow for an accurateassessment of the EAL.Core uncovery can be determined by the secondary indications listed in this EAL. The secondary indicators ofinventory loss include a list of tanks/sumps found in OPOP04-RC-0003, Excessive RCS Leakage. Since othersystem leaks could raise levels in various tanks and sumps, the list has been limited to the tanks and sumps thatwould 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 WaterLevels during Refueling for EAL Thresholds2. OPOP03-ZG-0009, Rev. 59, Mid-Loop Operation, Addendum 1, RCS/RHIR Simplified ElevationDiagram3. USFAR, Rev. 15, Chapter 12, Table 12.3.4-14. OPOP02-11-0002, Rev. 15, RVWL Monitoring System5. OPOP04-RC-0003, Rev 18, Excessive RCS Leakage6. Drawing 9C129A81105, Re. 3, Radiation Zones, Reactor Containment Building, Plan at E. 68' -0"59 1 Page CG1ECL: GENERAL EMERGENCYInitiating Condition: Loss of RCS inventory affecting fuiel clad integrity with containment challenged.Operating Mode Applicability: 5, 6Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the GENERAL EMERGENCY promptly upon determining that30 minutes has been exceeded, or will likely be exceeded.(1) a. RCS level less than 0% of plenum for 30 minutes or longer.ANDb. ANY indication from the Table C1.(2) a. RCS level cannot be monitored for 30 minutes or longer.ANDb. Core uncovery is indicated by ANY of the following:" Reactor Containment Building, 68'-0" Area Radiation Monitors RE-8055 or RE-8099 readinggreater than 9,000 mR/hr.OR" Erratic source range monitor indicationOR* UNPLANNED rise in ANY of the following sump or tank levels in Table C2 of sufficientmagnitude to indicate core uncoveryANDc. ANY indication from Table C1Table Cl: Containment Challenge" CONTAINMENT CLOSURE not established ** >4% hydrogen exists inside containment* UNPLANNED rise in containment pressure* IF CONTAINMENT CLOSURE is re-establishedprior to exceeding the 30-minute time limit, THENdeclaration 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 SumpBasis:This IC addresses the inability to restore and maintain RCS level above the top of active fuel with containmentchallenged. This condition represents actual or IMMINENT substantial core degradation or melting withpotential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposurelevels offsite for more than the immediate site area.Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactorcoolant 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 releaseof 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 hydrogenconcentration is sufficient to Support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn willraise containment pressure and could result in collateral equipment damage leading to a loss of containmentintegrity. 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 couldresult in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service duringan event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gasconcentration reading as ambient conditions within the containment will preclude personnel access. Duringperiods when installed containment hydrogen gas monitors are out-of-service, operators may use indications inTable 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 abilityto monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions todetermine if core uncovery has actually occurred (i.e., to account for various accident progression andinstrumentation 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 leveldropping below the range of available instrumentation. If water level cannot be monitored, operators maydetermine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tanklevel changes must be evaluated against other potential sources of water flow to ensure they are indicative ofleakage from the RCS61 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 atCommercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industrjy Actionsto 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 RCSlevel 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 MonitoringSystem thermocouple to TAF is Sensor 8 at elevation 29'-2.7". Use of RVWL to approximate TAF; with theinherent gap of 12 inches between indicated level and actual level, is acceptable for the purposes of maintainingthe 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, ExcessiveRCS Leakage. Since other system leaks could rise levels in various tanks and sumps, the list has been limited tothe tanks and sumps that would have the highest probability of indicating RCS leakage inside the ReactorContainment Building.As RCS level drops the dose rates above the core will rise. Area Radiation Monitors RE-8055 and RE-8099 arelocated on the 68'-0" elevation of the reactor containment building. Their locations are identified on drawing9C129A81105. Their range (0.1 mR/hr to 10,000 mR/hr) is identified in Table 12.3.4-1 of Section 12 of theUFSAR. 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 coreuncovery.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 fueland 6 R/hr at one foot above the top of active fuel. The high range of these monitors is 10 R/hr. The value of9,000 mR/hr was selected for this threshold to ensure the threshold is readily assessable and within the calibratedrange of the monitor. The threshold value of 9,000 lnR/hr with the reactor head on corresponds to approximately8 inches above the top of the active fuel which provides an additional indication that RCS levels are near thepoint of fuel uncovery. These monitor readings in conjunction with the other threshold values allow for anaccurate assessment of the EAL.REFERENCES:I. Calculation No. STPNOC013-CALC-006 Rev.2, Dose Rate Evaluation of Reactor Vessel WaterLevels during Refueling for EAL Thresholds2. OPOP03-ZG-0009, Rev. 59, Mid-Loop Operations3. 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 Monitors5. OPOP05-EO-E010, Rev. 21, Loss of Reactor or Secondary Coolant6. OPOP04-RC-0003, Rev. 18, Excessive RCS Leakage62 1 P a e 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)ICS/EALSTable E-1: Recognition Category "E" Initiating Condition MatrixUNUSUAL EVENTE-HU1 Damage to a loaded caskCONFINEMENT BOUNDARY.Op. Modes: ALL631 P a ge E-HU1ECL: UNUSUAL EVENTInitiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARYOperating Mode Applicability: ALLEmergency Action Level:(1) Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiationreading greater than: :a. 60 mrem/hr (gamma + neutron) on the top surface of the spent fuel caskORb. 600 mreni/hr (gammna + neutron) onl the side surface of the spent fuel caskORb. 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 caskcontaining spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that theloaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to theenvironment, degradation of one or more fuel assemblies due to environmental factors, and configurationchanges 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 theTechnical Specification allowable radiation levels. The technical specification multiple of "2 times", which isalso used in Recognition Category R IC RUI. is used here to distinguish between non-emergency and emergencyconditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask andnot the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to aloaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurementof 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. Section5.3.2 of the "Certificate of Compliance No. 1032, Appendix A, Technical Specifications For The HI-STORMFW MPC Storage System", states:5.3.4 Notwithstandizg the limits established in Section 5.3.3, the measured closerates on a loaded OVERPACK or TRANSFER CASK shall not exceed theJbllowing values:a. 30 mrem/hr (gamma + neutron) on the top qf the OVERPACIC641 P a e
- b. 300 mrem/hr (gamma + neutron) on the side of the OVERPACK.excluding inlet and outlet ductsc. 3500 mrem/hr (gamma + neutron) on the side of the TRANSFERCASKREFERENCES:1. Certificate of Compliance no. 1032, Appendix A, Technical Specifications For The HI-STORM FWMPC Storage System, Section 5.3, Radiation Protection Program.10 CFR 72.104, Criteria ForRadioactive Materials In Effluents And Direct Radiation From An ISFSI or MRS65 1 P a g e 9 FISSION PRODUCT BARRIER ICS/EALSTable 9-F-I: Recognition Category "F" Initiating Condition MatrixALERTFA1 ANY Loss or ANY Potential Loss of either theFuel Clad or RCS barrier.tOp. Modes: 1,2,3,4SITE AREA EMERGENCYFS1 Loss or Potential Loss of ANY two barriers.Op. Modes: 1,2,3,4GENERAL EMERGENCYFG1 Loss of ANY two barriers and Loss or PotentialLoss of the third barrier.Op. Modes: 1,2,3,4661 P ag e Table 9-F-2: EAL Fission Product Barrier TableThresholds for LOSS or POTENTIAL LOSS of BarriersFA1 ALERT FS1 SITE AREA EMERGENCY FG1 GENERAL EMERGENCYANY 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 PotentialClad or RCS barrier. Loss of the third barrier.Fuel Clad Barrier RCS Barrier Containment BarrierLOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS1. RCS or SG Tube Leakage 1. RCS or SG Tube Leakage 1. RCS or SG Tube LeakageNot Applicable A. Core Cooling -Orange A. An automatic or A. Operation of a standby A. A leaking or Not Applicableentry conditions met manual ECCS (SI) charging pump is RUPTURED SGactuation is required required by EITHER is FAULTEDby EITHER of the of the following: outside offollowing: containment.1. UNISOLABLE1. UNISOLABLE RCS leakageRCS leakage OROR 2. SG tube leakage.2. SG tubeRUPTURE.B. Integrity -Red entryconditions met67 1 P a g e Fuel Clad Barrier RCS Barrier Containment BarrierLOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS2. Inadequate Heat Removal 2. Inadequate Heat Removal 2. Inadequate Heat RemovalA. Core Cooling -Red A. Core Cooling -Not Applicable A. Heat Sink -Red Not Applicable A. Core Cooling- Redentry conditions met Orange entry entry conditions entry conditions metconditions met met. for 15 minutes orOR longer.B. Heat Sink- Redentry conditions met3. RCS Activity / Containment Radiation 3. RCS Activity / Containment Radiation 3. RCS Activity / Containment RadiationAl. RCB Rad Monitor Not Applicable A. Not Applicable Not Applicable Not Applicable Al. RCB Rad MonitorRT-8050 or RT- RT-8050 or RT-8051 greater than 8051 greater than40 R/hr 380 R/hrOR OR2. HATCH 2. HATCHMONITOR MONITORgreater than 90 greater than 840mR/hr mR/hrORB. Sample analysisindicates that reactorcoolant activity isgreater than 300jiCi/gm doseequivalent 1- 13 1.68 1 P a g e Fuel Clad Barrier RCS Barrier Containment BarrierLOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS4. Containment Integrity or Bypass 4. Containment Integrity or Bypass 4. Containment Integrity or BypassNot Applicable Not Applicable Not Applicable Not Applicable A. Containment A. Containment -Redisolation is required entry conditions metAND EITHER of ORthe following: B. Explosive mixture1. Containment exists insideintegrity has containmentbeen lost based (H2 > 4%)on Emergency ORDirector CI. Containmentjudgment. pressure greaterOR than 9.5 psig.AND2. UNISOLABLEpathway from 2. Less than one fullthe containment train ofto the Containment Sprayenvironment is operating perexists. design for 15minutes or longer.ORB. Indications of RCSleakage outside ofcontainment.69 I P a g e Fuel Clad Barrier RCS Barrier Containment BarrierLOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS5. Other Indications 5. Other Indications 5. Other IndicationsA. N/A A. N/A A. N/A A. N/A A. N/A A. N/A6. Emergency Director Judgment 6. Emergency Director Judgment 6. Emergency Director JudgmentA. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition inthe opinion of the the opinion of the the opinion of the the opinion of the the opinion of the the opinion of theEmergency Director Emergency Director Emergency Director Emergency Director Emergency Director Emergency Directorthat indicates Loss that indicates that indicates Loss that indicates that indicates Loss that indicatesof the Fuel Clad Potential Loss of the of the RCS Barrier. Potential Loss of the of the Containment Potential Loss of theBarrier. Fuel Clad Barrier. RCS Barrier. Barrier. ContainmentBarrier.70 1 P a g e Basis Information ForEAL Fission Product Barrier Table 9-F-2STP is part of the Westinghouse Owners Group (WOG) and has adopted the WOG Emergency ResponseGuidelines (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 CSFSTassessment results as EALs and fission product barrier loss/potential loss thresholds. This approachallows consistency between EOPs and emergency classifications.711 P a e FUEL CLAD BARRIER THRESHOLDSThe Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.I1. RCS or SG Tube LeakageLoss IThere is no Loss threshold associated with RCS or SG Tube LeakageLPotential Loss 1.ACore Cooling -Orange entry conditions (CETs > 7080 F) are sufficient to allow the onset of heat-inducedcladding damage.2. Inadequate Heat RemovalLoss 2.ACore Cooling -Red entry conditions (CETs > 12000 F) are sufficient to cause significant superheating ofreactor coolant.Potential Loss 2.ACore Cooling -Orange entry conditions (CETs > 708' F) are sufficient to allow the onset of heat-inducedcladding damage.Potential Loss 2.BHeat 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 thesteam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potentialloss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions duringwhich operators intentionally reduce the heat removal capability of the steam generators; during theseconditions, classification using threshold is not warranted.Meeting this threshold results in a SITE AREA EMERGENCY because this threshold is identical to RCSBarrier Potential Loss threshold 2.A; both will be met. This condition warrants a SITE AREAEMERGENCY declaration because inadequate RCS heat removal may result in fuel heat-up sufficient todamage 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 THRESHOLDS3. RCS Activity / Containment RadiationLoss 3.A. 1The readings for the containment high range area monitors (RT-8050 and RT-805 1) correspond to aninstantaneous release of all reactor coolant mass into the containment, assuming that reactor coolantactivity equals 300ýiCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater thanthat 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 aloss of the Fuel Clad Barrier. The values for RT-8050 and RT-8051 were based on CalculationSTPNOC013-004 Rev.2. The threshold values were conservatively rounded within 2% of the calculatedvalues to make the values readily assessable. Temperature induced current (TIC) limitations are notapplicable to the Fuel Clad Barrier Loss threshold 3.A. 1 because the expected radiation dose for this eventoverwhelms the TIC effect. This effect is discussed in the IOCFR50.59 evaluation 04-8245-60 associatedwith DCP 04-8245-3 3.Loss 3.A.2The HATCH MONITOR is located outside containment and is the back-up monitor to the containmenthigh range monitors (RT-8050 and RT-805 1). The HATCH MONITOR threshold value is based onCalculation No. 03-ZE-003. This value corresponds to the calculated containment high range monitorreadings for Fuel Clad Barrier Loss 3.AThe radiation monitor reading in this threshold is higher than that specified for RCS Barrier Lossthreshold 3.A since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that acombination of the two monitor readings appropriately escalates the EMERGENCY CLASSIFICATIONLEVEL to a SITE AREA EMERGENCY.Loss 3.BThis threshold indicates that RCS radioactivity concentration is greater than 300 [LCi/gm doseequivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodinespikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this conditionindicates that a significant amount of fuel clad damage has occurred, it represents a loss of the FuelClad Barrier.Potential Loss 3.There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.4. Containment Integrity or BypassNot Applicable (included for numbering consistency)731 P a -e FUEL CLAD BARRIER THRESHOLDS5. Other IndicationsLoss and/or Potential Loss 5.AN/A6. Emergency Director JudgmentLoss 6.AThis threshold addresses any other factors that may be used by the Emergency Director in determiningwhether the Fuel Clad Barrier is lost.Potential Loss 6.AThis threshold addresses any other factors that may be used by the Emergency Director in determiningwhether the Fuel Clad Barrier is potentially lost. The Emergency Director should also consider whether ornot to declare the barrier potentially lost in the event that barrier status cannot be monitored.741 P a o e RCS BARRIER THRESHOLDSThe RCS Barrier includes the RCS primary side and its connections Lip to and including the pressurizersafety and relief valves, and other connections tip to and including the primary isolation valves.1. RCS or SG Tube LeakageLoss 1.AThis threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic ormanual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents aloss 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 beinto any location -inside containment, to the secondary-side (i.e., steam generator tube leakage) oroutside of containment.A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injectionis considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside ofcontainment, the declaration escalates to a SITE AREA EMERGENCY since the Containment BarrierLoss threshold I.A will also be met.Potential Loss I.AThis threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizerlevel 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 crewsupervision, directs that a standby charging (makeup) pump be placed in service to restore and maintainpressurizer 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 beinto any location -inside containment, to the secondary-side (i.e., steam generator tube leakage) oroutside of containment.If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a SITEAREA EMERGENCY since the Containment Barrier Loss threshold L.A will also be met.Potential Loss I.BIntegrity -Red entry conditions indicate an extreme challenge to the integrity of the RCS pressureboundary due to pressurized thermal shock -a transient that causes rapid RCS cooldown while the RCSis in Mode 3 or higher (i.e., hot and pressurized).75 1 P a -e RCS BARRIER THRESHOLDS2. Inadequate Heat RemovalLoss 2.AThere is no Loss threshold associated with Inadequate Heat Removal.Potential Loss 2.AHeat Sink- Red entry conditions met (NR level in all SG < 14% [34%] AND total AFW flow toSGs < 576 GPM).This condition indicates an extreme challenge to the ability to remove RCS heat using the steamgenerators (i.e., loss of an effective secondary-side heat sink). This condition represents a potentialloss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions duringwhich operators intentionally reduce the heat removal capability of the steam generators; duringthese conditions, classification using threshold is not warranted.Meeting this threshold results in a SITE AREA EMERGENCY because this threshold is identical to FuelClad Barrier Potential Loss threshold 2.B; both will be met. This condition warrants a SITE AREAEMERGENCY declaration because inadequate RCS heat removal may result in fuel heat-up sufficient todamage the cladding and raise RCS pressure to the point where mass will be lost from the system.3. RCS Activity / Containment RadiationLoss 3.A.Not ApplicablePotential Loss 3.There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.4. Containment Integrity or BypassNot Applicable (included for numbering consistency)761 Pag e RCS BARRIER THRESHOLDS5. Other IndicationsLoss and/or Potential Loss 5.AVariables 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 responsecharacteristics of Type C information display channels allow the control room operator to detectconditions indicative of significant failure of any of the three fission product barriers or the potentialfor significant failure of these barriers. Although variables selected to fulfill Type C functions mayrapidly approach the values that indicate an actual significant failure, it is the final steady-state valuereached that is important. Therefore, a high degree of accuracy and a rapid response time are notnecessary for Type C information display channels. Type C variables are found in UFSAR Table7B.6- 1.6. Emergency Director JudgmentLoss 6.AThis threshold addresses any other factors that may be used by the Emergency Director in detenniningwhether the RCS Barrier is lost.Potential Loss 6.AThis threshold addresses any other factors that may be used by the Emergency Director in determiningwhether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or notto declare the barrier potentially lost in the event that barrier status cannot be monitored.77 1 P a e CONTAINMENT BARRIER THRESHOLDSTile Containment Barrier includes the containment building and connections uip to and including theoutermost containment isolation valves. This barrier also includes the main steam, feedwater, andblowdown line extensions outside the containment building uip to and including the outermost secondaryside isolation valve. Containment Barrier thresholds are used as criteria for escalation of the ECL fromALERT to a SITE AREA EMERGENCY or a GENERAL EMERGENCY.l. RCS or SG Tube LeakageLoss L.AThis threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outsideof containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordancewith the thresholds for RCS Barrier Potential Loss 1.A and Loss 1.A, respectively. This conditionrepresents a bypass of the containment barrier.FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarilydependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steamgenerator is decreasing uncontrollably [part of the FAULTED definition] and the FAULTED steamgenerator isolation procedure is not entered because EOP user rules are dictating implementation ofanother procedure to address a higher priority condition, the steam generator is still consideredFAULTED for emergency classification purposes.The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that mayrequire an emergency classification. Steam releases of this size are readily observable with normalControl Room indications. The lower bound for this aspect of the containment barrier is analogous to thelower bound criteria specified in IC SU3 for the fuel clad barrier (i.e., RCS activity values) and IC SU4for the RCS barrier (i.e., RCS leak rate values).This threshold also applies to prolonged steam releases necessitated by operational considerations such asthe forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown theplant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in asignificant and sustained release of radioactive steam to the environment (and are thus similar to aFAULTED condition). The inability to isolate the steam flow without an adverse effect on plantcooldown meets the intent of a loss of containment.Steam releases associated with the expected operation of a SG power operated relief valve or safety reliefvalve do not meet the intent of this threshold. Such releases may occur intermittently for a short period oftime following a reactor trip as operators process through emergency operating procedures to bring theplant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with theunexpected 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-sidesystem component (e.g., air ejectors., glad seal exhausters, valve packing, etc.). These types of releases donot constitute a loss or potential loss of containment but should be evaluated using the RecognitionCategory R ICs.781 Pa- e CONTAINMENT BARRIER THRESHOLDSThe EMERGENCY CLASSIFICATION LEVELS resulting from primary-to-secondary leakage, with orwithout a steam release from the FAULTED SG, are summarized below.Affected SG is FAULTED Outside of Containment?P-to-S Leak RateYesNoLess than or equal to 25gpmGreater than 25 gpmRequires operation of astandby charging pump(RCS Barrier PotentialLoss)Requires an automatic ormanual ECCS (SI)actuation (RCS BarrierLoss)No classificationUNUSUAL EVENT perSU4SITE AREA EMERGENCYper FS ISITE AREA EMERGENCYper FS INo classificationUNUSUAL EVENT perSU4ALERT per FAlALERT per FAIPotential Loss 1.There is no Potential Loss threshold associated with RCS or SG Tube Leakage.2. Inadequate Heat RemovalLoss 2There is no Loss threshold associated with Inadequate Heat Removal.Potential Loss 2.ACore Cooling -Red entry conditions met for 15 minutes or longer. This condition represents anIMMINENT core melt sequence which, if not corrected, could lead to vessel failure and a higher potentialfor containment failure. For this condition to occur there must already have been a loss of the RCS Barrierand the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is noteffective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to coremelting and a subsequent challenge of the Containment Barrier.The restoration procedure is considered "effective" if core exit thermocouple readings are decreasingand/or if RCS level is increasing. Whether or not the procedure(s) will be effective should be apparentwithin 15 minutes. The Emergency Director should escalate the emergency classification level as soon asit is determined that the procedure(s) will not be effective.Severe accident analyses (e.g., NUREG-1 150) have concluded that function restoration procedures canarrest core degradation in a significant fraction of core damage scenarios, and that the likelihood ofcontainment failure is very small in these events. Given this, it is appropriate to provide 15 minutesbeyond the required entry point to determine if procedural actions can reverse the core melt sequence.791 P a g e CONTAINMENT BARRIER THRESHOLDS3. RCS Activity / Containment RadiationLoss 3There is no Loss threshold associated with RCS Activity / Containment Radiation.Potential Loss 3.A.IThe readings for the containment high range area monitors (RT-8050 and RT-805 1) correspond to aninstantaneous release of the radioactive material inventory of the reactor coolant system (i.e., All the RCScoolant 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 usedwere conservatively rounded within 2% of the calculated values to ensure the values were readilyassessable. This level of assumed fuel clad failure is well beyond that used to determine the analogousFuel Clad Barrier Loss and RCS Barrier Loss thresholds. Temperature induced current (TIC) limitationsare not applicable to the Containment Barrier Potential Loss threshold 3.A.I because the expectedradiation dose for this event overwhelms the TIC effect. This effect is discussed in 1OCFR50.59evaluation 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 majorrelease of radioactivity requiring offsite protective actions. For this condition to exist, there must alreadyhave been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat thiscondition as a potential loss of containment which would then escalate the EMERGENCYCLASSIFICATION LEVEL to a GENERAL EMERGENCY.Potential Loss 3.A.2The HATCH MONITOR is located outside containment and is the back-up monitor to the containmenthigh range monitors (RT-8050 and RT-805 1). The HATCH MONITOR threshold value is based onCalculation No. 03-ZE-003. This value corresponds to the calculated containment high range monitorreadings for Containment Barrier Threshold Potential Loss 3.A. 1.4. Containment Integrity or BypassLoss 4.AThese thresholds address a situation where containment isolation is required and one of two conditionsexists as discussed below. Users are reminded that there may be accident and release conditions thatsimultaneously 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 likelyexceeds that associated with allowable leakage (or sometimes referred to as design leakage). Followingthe release of RCS mass into containment., containment pressure will fluctuate based on a variety offactors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable dropin containment pressure. Recognizing the inherent difficulties in determining a containment leak rate80 1 P a , e CONTAINMENT BARRIER THRESHOLDSduring accident conditions, it is expected that tile Emergency Director will assess this threshold usingjudgment, and with due consideration given to current plant conditions, and available operational andradiological 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 leakagefrom a penetration and the other is leakage from an in-service system valve. Depending upon radiationmonitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted inthe figure.Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence oftwo FAULTED locations on a steam generator where one fault is located inside containment (e.g., on asteam or feedwater line) and the other outside of containment. In this case, the associated steam lineprovides 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 beminor radiological releases associated with allowable (design) containment leakage through variouspenetrations or system components. These releases do not constitute a loss or potential loss ofcontainment 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 radioactivematerial 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 withthe 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 dropin 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 containmentisolation was required (i.e., containment isolation was not successful). There is now an UNISOLABLEpathway from the containment to the enviromnent.The existence of a filter is not considered in the threshold assessment. Filters do not remove fissionproduct noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loadingbeyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/highhumidity 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 cooleris allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected bythe Process Monitor. If there is no leakage from the Component Cooling Water system to the AuxiliaryBuilding, then no threshold has been met. If the pump or system piping developed a leak that allowedsteam/water to enter the Auxiliary Building, then threshold 4.B would be met. Depending upon radiationmonitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted inthe figure and cause threshold 4.A.1 to be met as well.81 Page CONTAINMENT BARRIER THRESHOLDSFollowing the leakage of RCS mass into containment and a rise in containment pressure, there may beminor radiological releases associated with allowable (design) containment leakage through variouspenetrations 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 constitutea 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 assessedusing Loss Threshold I.A.Loss 4.BContainment sump, temperature, pressure and/or radiation levels will rise if reactor coolant mass isleaking into the containment. If these parameters have not risen, then the reactor coolant mass may beleaking 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 massis being lost outside of containment.Unexpected elevated readings and alarms on radiation monitors with detectors outside containmentshould be corroborated with other available indications to confirm that the source is a loss of RCS massoutside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside ofcontainment may not rise significantly; however, other unexpected changes in sump levels, areatemperatures or pressures, flow rates, etc. should be sufficient to determine if RCS mass is being lostoutside of the containment.Refer to the middle piping run of Figure 9-F-3. In this simplified example, a leak has occurred at areducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitorlocations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figureand cause threshold 4.A.I to be met as well.To ensure proper escalation of the emergency classification, the RCS leakage outside of containmentmust be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold I.A to bemet.Potential Loss 4.AContainment -Red entry conditions met (containment pressure > 56.5 PSIG). If containment pressureexceeds 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 andFuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a SITE AREAEMERGENCY and GENERAL EMERGENCY since there is now a potential to lose the third barrier.Potential Loss 4.BThe existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogenconcentration is sufficient to support a hydrogen bum (i.e., at the lower deflagration limit (4%)). Ahydrogen burn will raise containment pressure and could result in collateral equipment damage leading toa loss of containment integrity. It therefore represents a potential loss of the Containment Barrier.82 1 P a g e CONTAINMENT BARRIER THRESHOLDSPotential Loss 4.CThis threshold describes a condition where containment pressure is greater than the setpoint (9.5 PSIG) atwhich Contaimnent Spray is designed to automatically actuate, and less than one full train of equipment iscapable of operating per design. The 15-minute criterion is included to allow operators time to manuallystart equipment that may not have automatically started, if possible. This threshold represents a potentialloss of containment in that Containment Spray is either lost or performing in a degraded manner.5. Other IndicationsLoss and/or Potential Loss 5.AN/A6. Emergency Director JudgmentLoss 6.AThis threshold addresses any other factors that may be used by the Emergency Director in determiningwhether the Containment Barrier is lost.Potential Loss 6.AThis threshold addresses any other factors that may be used by the Emergency Director in determiningwhether the Containment Barrier is potentially lost. The Emergency Director should also considerwhether 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 ExamplesRCP SealCoolingNOTES: Only Supplemental Purge is a filtered release and STPEGS Component Cooling Water is equivalent to Closed Cooling Water841 P ag e 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANTSAFETY ICS/EALSTable H-I: Recognition Category "H" Initiating Condition MatrixUNUSUAL EVENTHU1 ConfirmedSECURITY CONDITIONor threat.Op. Modes: ALLALERTSITE AREAEMERGENCYGENERALEMERGENCYHU2 Seismic event greaterthan OBE levels.Op. Modes: ALLHU3 Hazardous event.Op. Modes: ALLHU4 FIRE potentiallydegrading the level ofsafety of the plant.Op. Modes: ALLHA1 HOSTILE ACTIONwithin the OWNERCONTROLLED AREA orairborne attack threatwithin 30 minutes.Op. Modes: ALLNote:See SA9 or CA6for escalation ofthese eventsHA5 Gaseous releaseimpeding access toequipment necessary fornormal plant operations,cooldown or shutdown.Op. Modes: ALLHA6 Control Roomevacuation resulting intransfer of plant control toalternate locations. Op.AModes: ALLHA7 Other conditionsexist which in thejudgment of theEmergency Directorwarrant declaration of anALERT.Op. Modes: ALLHS1 HOSTILE ACTIONwithin the PROTECTEDAREA.Op. Modes: ALLHG1 HOSTILE ACTIONresulting in loss of physicalcontrol of the facility.Op. Modes: ALLHU7 Other conditions existwhich in the judgment ofthe Emergency Directorwarrant declaration of anUNUSUAL EVENT.Op. Modes: ALLHS6 Inability to control akey safety function fromoutside the Control Room.Op. Modes: ALLHS7 Other conditions existwhich in the judgment ofthe Emergency Directorwarrant declaration of aSITE AREAEMERGENCY.Op. Alodes: ALLHG7 Other conditionsexist which in thejudgment of theEmergency Directorwarrant declaration of aGENERALEMERGENCY.Op. AMlodes: ALL85 I P a- e HU1ECL: UNUSUAL EVENTInitiating Condition: Confirmed SECURITY CONDITION or threat.Operating Mode Applicability: ALLEmergency Action Levels: (1 or 2 or 3)(1) A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by ANY ofthe 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 thusrepresent a potential degradation in the level of plant safety. SECURITY EVENTS which do not meetone 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 isessential for proper classification of a security-related event. Classification of these events will initiateappropriate threat-related notifications to plant personnel and OROs.Security plans and terminology are based on the guidance provided by NEI 03-12, Template./br theSecurity Plan, Training and Qualification Plan, Sa feguards Contingency Plan [and INDEPENDENTSPENT FUEL STORAGE INSTALLATION Security Program].EAL #1- references Security Force Supervisor because these are the individuals trained to confirm that aSECURITY EVENT is occurring or has occurred. Training on SECURITY EVENT confirmation andclassification 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 isassessed in accordance with OSDPOI-ZS-0011, Implementing Procedure For Safeguards ContingencyEvents.EAL #3- addresses the threat from the impact of an aircraft on the plant. The NRC HeadquartersOperations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The statusand size of the plane may also be provided by NORAD through the NRC. Validation of the threat is861Pag e performed in accordance with OPOP04-ZO-SEC4. Guideline For Airborne (Aircraft) Threat, and SecurityForce Instruction SI 2700, Security Response to Airborne Threat.Emergency plans and implementing procedures are public documents; therefore, EALs do not incorporateSecurity-sensitive information. This includes information that may be advantageous to a potentialadversary, such as the particulars concerning a specific threat or threat location. Security-sensitiveinformation 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 becausethe Acting Security Manager position is not normally activated until after an UNUSUAL EVENT hasbeen declared, and the Security Manager position is not normally activated until after an ALERT has beendeclared. However, reporting by the Acting Security Manager or Security Manager was included in theevent these positions are staffed under unusual circumstances.REFERENCEs:1. OERPOI-ZV-SH03, Rev. 12, Acting Security Manager2. OERPO1-ZV-TS08, Rev. 16, Security Manager3. 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 HU2ECL: UNUSUAL EVENTInitiating Condition: Seismic event greater than OBE levels.Operating Mode Applicability: ALLEmergency Action Level:(1) a. EITHER of the following conditions exist:1. "SEISMIC EVENT" alarm in Unit I Control Room (Lampbox 9M0I, Window E-8)OR2. Control Room personnel feel an actual or potential seismic event.ANDb. The occurrence of a seismic event is confirmed in manner deemed appropriate by the ShiftManager or Emergency Director.Basis:This IC addresses a seismic event that results in accelerations at the plant site greater than those specifiedfor an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a SafeShutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures andcomponents; however, some time may be required for the plant staff to ascertain the actual post-eventcondition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessaryto perform walk-downs and inspections, and fully understand any impacts, this event represents apotential 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), itdoes 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, SeismicEvent 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 ControlRoom personnel and which can be readily assessed. Indications available outside the Control Roomand/or which require lengthy times to assess (e.g., processing of scratch plates or recorded data) shouldnot be used. The goal is to specify indications that can be assessed within 15-minutes of the actual orsuspected seismic event.The EAL 1.b- statement is included to ensure that a declaration does not result from felt vibrations causedby a non-seismic source (e.g., a dropped heavy load). The Shift Manager or Emergency Director mayseek 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 recognizedthat 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 EMERGENCYCLASSIFICATION 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 hasexperienced an OBE. The Seismic Event Alarm setpoint is 0.02g in the vertical or horizontal position andthe station design basis value for an OBE is 0.05g. Since the Seismic Event alarm is set at less than halfof the OBE value, it cannot be used as the sole threshold value for detennining whether or not STP hasexperienced an OBE.STP has implemented the alternative EAL described in NEI 99-01 Developer Notes in conjunction withusing the installed indication. EAL-1, b. allows the Shift Manager or Emergency Director to determine ifa seismic event has taken place, taking into consideration the Seismic Event alarm, Control Roompersonnel feeling an actual or potential seismic event and other indications deemed appropriate.REFERENCES:1. OPOP04-SY-0001, Rev. 8, Seismic Event2. NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors.89 1 P a ge HU3ECL: UNUSUAL EVENTInitiating Condition: Hazardous event.Operating Mode Applicability: ALLEmergency 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 vehiclebreakdowns 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 electricalisolation 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 eventinvolving 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 fromaccessing the site via personal vehicles.(5) Predicted or actual breach of Main Cooling Reservoir retaining dike along North WallBasis:This IC addresses hazardous events that are considered to represent a potential degradation of the level ofsafety 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 aSAFETY SYSTEM component due to water level or other wetting concerns. Classification is alsorequired if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEMcomponent from its power source (e.g., a breaker or relay trip). To warrant classification, operability ofthe 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 sufficientmagnitude 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 andsignificant enough to prohibit the plant staff fi'om accessing the site using personal vehicles. Examples ofsuch an event include site flooding caused by a hurricane, heavy rains, up-river water releases, damfailure, etc., or an on-site train derailment blocking the access road. This EAL is not intended apply toroutine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to moresignificant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around90 I P a -e the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in2011.EAL#5- the Main Cooling Reservoir breach along the north wall which was included because it is acredible hazard and analyzed in the STPEGS UFSAR.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be based on ICs in RecognitionCategories R, F, S or C.HU3: EAL-1, EAL-2, EAL-3, EAL-4 Selection Basis:N/AREFERENCE:1. STPEGS UFSAR, Section 3.4.1, Flood Protection911 1 a " e HU4ECL: UNUSUAL EVENTInitiating Condition: FIRE potentially degrading the level of safety of the plant.Operating Mode Applicability: ALLEmergency Action Levels: (1 or 2 or 3 or 4)Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining thatthe 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 detectionindications:* Report from the field (i.e., visual observation)* Receipt of multiple (more than 1) fire alarms or indications* Field verification of a single fire alarmANDb. 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).ANDb. The FIRE is located within ANY of the plant rooms or areas in Table H4:ANDc. 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 ofthe initial report, alarm or indication.(4) A FIRE within the ISFS1 OR plant PROTECTED AREA that requires firefighting support by anoffsite 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 ofthe level of safety of the plant.EAL #1The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that arereadily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of aFIRE 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 initialalarm, indication, or report was received, and not the time that a subsequent verification action wasperformed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm,indication or report.EAL #2This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., provedor disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirmthe validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the timethat 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 aspurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify thevalidity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIREexists; however, after that time, and absent information to the contrary, it is assumed that an actual FIREis in progress.If an actual FIRE is verified by a report from the field, then EAL #1 is immediately applicable, and theemergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarmis verified to be due to an equipment failure or a spurious activation, and this verification occurs within30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration iswarranted.EAL #3In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant or ISFSI PROTECTEDAREA not extinguished within 60-minutes may also potentially degrade the level of plant safety.EAL #4If a FIRE within the plant or ISFSI PROTECTED AREA is of sufficient size to require a response by anoffsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentiallydegraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration onlyif it is needed to actively support firefighting efforts because the fire is beyond the capability of the FireBrigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, orsupporting post-extinguishment recovery or investigation actions.93 1P a g e Basis-Related Reouirements from Annendix RAppendix R to 10 CFR 50, states in part:Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components importantto safety shall be designed and located to minimize, consistent with other safety requirements, theprobability and effect of fires and EXPLOSIONS."When considering the effects of fire, those systems associated with achieving and maintaining safeshutdown conditions assume major importance to safety because damage to them can lead to coredamage resulting from loss of coolant through boil-off.Because fire may affect safe shutdown systems and because the loss of function of systems used tomitigate the consequences of design basis accidents under post-fire conditions does not per se impactpublic safety, the need to limit fire damage to systems required to achieve and maintain safeshutdown conditions is greater than the need to limit fire damage to those systems required tomitigate the consequences of design basis accidents.In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour firebarriers 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-hourtime period.Depending upon the plant mode at the time of the event, escalation of the EMERGENCYCLASSIFICATION 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 Program2. STPEGS UFSAR, Rev. 16, Section 7.4, Systems Required for Safe Shutdown94 P a g e HU7ECL: UNUSUAL EVENTInitiating Condition: Other conditions exist which in the judgment of the Emergency Director warrantdeclaration of a UE.Operating Mode Applicability: ALLEmergency Action Level:(1) Other conditions exist which in the judgment of the Emergency Director indicate that events arein progress or have occurred which indicate a potential degradation of the level of safety of theplant or indicate a security threat to FACILITY protection has been initiated. No releases ofradioactive material requiring offsite response or monitoring are expected unless furtherdegradation of safety systems occurs.Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declarationof an emergency because conditions exist which are believed by the Emergency Director to fall under theEMERGENCY CLASSIFICATION LEVEL description for an UE.HU7: EAL-1 Selection Basis:N/AREFERENCES:N/A95 1 P a o e HA1ECL: ALERTInitiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborneattack threat within 30 minutes.Operating Mode Applicability: ALLEmergency Action Levels: (1 or 2)(1) A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREAas 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 AREAor notification of an aircraft attack threat. This event will require rapid response and assistance due to thepossibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and stafffor a potential aircraft impact.Timely and accurate communications between Security Shift Supervision and the Control Room isessential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for theSecuritv Plan. Training and Qualification Plan, Safeguards Contingency Plan [and INDEPENDENTSPENT FUEL STORAGE INSTALLA TION Security Program].As time and conditions allow, these events require a heightened state of readiness by the plant staff andimplementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The ALERTdeclaration will also heighten the awareness of Offsite Response Organizations, allowing them to bebetter 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 arenot a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a smallaircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of eventsis 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 OWNERCONTROLLED AREA.961Pagc EAL #2 addresses the threat friom tile impact of an aircraft on the plant, and the anticipated arrival time iswithin 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in atimely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is metwhen the threat-related information has been validated in accordance with OPOP04-ZO-SEC4, Guidelinesfor Airborne (Aircraft) Threat, and Security Force Instruction SI 2700, Security Response to AirborneThreat.The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involvesan 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 CONTROLLEDAREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notificationby an appropriate Federal agency to tile site would clarify this point. In this case, the appropriate federalagency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one basedon 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 incorporateSecurity-sensitive information. This includes information that may be advantageous to a potentialadversary, such as the particulars concerning a specific threat or threat location. Security-sensitiveinfonnation 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 ORActing Security Manager were included because either of these positions could be activated prior tomeeting this EAL. The Security Force Supervisor is a 24-hour position and the normally the ActingSecurity Manager is activated after an UNUSUAL EVENT has been declared. The Security Manager isalso included although this position is normally activated after an ALERT.REFERENCES:1. OERPO1-ZV-SH03, Rev. 12, Acting Security Manager2. OERPO1-ZV-TS08, Rev. 16, Security Manager3. 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 HA5ECL: ALERTInitiating Condition: Gaseous release impeding access to equipment necessary for normal plantoperations, cooldown or shutdown.Operating Mode Applicability: ALLEmergency Action Level:Note: If the equipment in the listed room or area was already inoperable or out-of-service before theevent 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 theplant rooms or areas listed in Table H3/R2:ANDb. Entry into the room or area is prohibited or impeded.TABLE H3/R2: Plant Areas Requiring AccessRCB RHR Heat Exchanger Rooms0C ) MAB 51 ft Room 335EAB Roof, MCC 1G8, 4.16KV Switchgear Rooms00 Ln EAB 4.16KV Switchgear Rooms98 1P a , e Basis:This IC addresses an event involving a release of a hazardous gas that precludes or impedes access toequipment necessary to maintain normal plant operation, or required for a normal plant cooldown andshutdown. This condition represents an actual or potential substantial degradation of the level of safety ofthe plant.An ALERT declaration is warranted if entry into the affected room/area is, or may be, procedurallyrequired during the plant operating mode in effect at the time of the gaseous release. The emergencyclassification 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 EmergencyDirector's judgment that the gas concentration in the affected room/area is sufficient to preclude orsignificantly impede procedurally required access. This judgment may be based on a variety of factorsincluding an existing job hazard analysis, report of ill effects on personnel, advice from a subject matterexpert or operating experience with the same or similar hazards. Access should be considered as impededif 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 fornormal operation, cooldown and shutdown do not require entry into the affected room until Mode4." The gas release is a planned activity that includes compensatory measures which address thetemporary 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 actuallyprevent or impede a required action.An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Mostcommonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces theconcentration 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 firesuppression system in an area.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via Recognition Category R, Cor 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 safeshutdown equipment operation is capable from the Main Control Room (MCR). The plant is able totransition 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 necessaryfollowing entry into hot shutdown (establish Residual Heat Removal shutdown cooling, disable operationof 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 willbe maintained in a cold shutdown condition.REFERENCES:1. OPGP03-ZF-0001, Rev. 26, Fire Protection Program2. STPEGS UFSAR, Rev. 16, Section 7.4, Systems Required for Safe Shutdown3. 0POP03-ZG-0008, Rev. 56, Power Operations4. OPOP03-ZG-0006, Rev. 54, Plant Shutdown from 100% to Hot Standby5. OPOP03-ZG-0007, Rev. 71, Plant Cooldown100 P ag e HA6ECL: ALERTInitiating Condition: Control Room evacuation resulting in transfer of plant control to alternatelocations.Operating Mode Applicability: ALLEmergency Action Level:(1) An event has resulted in plant control being transferred firom the Control Room to the AuxiliaryShutdown Panel (ASP).Basis:This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternatelocations outside the Control Room. The loss of the ability to control the plant from the Control Room isconsidered 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 shutdownlocations. The necessity to control a plant shutdown from outside the Control Room, in addition toresponding to the event that required the evacuation of the Control Room, will present challenges to plantoperators and other on-shift persormel. Activation of the ERO and emergency response facilities willassist 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, asthe location where plant control is transferred in the event of a Control Room evacuation.REFERENCES:I. Procedure OPOP04-ZO-0001, Rev. 35, Control Room Evacuation101 lPage HA7ECL: ALERTInitiating Condition: Other conditions exist which in the judgment of the Emergency Directorwarrant declaration of an ALERT.Operating Mode Applicability: ALLEmergency Action Level:(1) Other conditions exist which, in the judgment of the Emergency Director, indicate that events arein progress or have occurred which involve an actual or potential substantial degradation of thelevel of safety of the plant or a SECURITY EVENT that involves probable life threatening risk tosite personnel or damage to site equipment because of HOSTILE ACTION. ANY releases areexpected to be limited to small fractions of the EPA PROTECTIVE ACTION GUIDELINEexposure levels.Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declarationof an emergency because conditions exist which are believed by the Emergency Director to fall under theEMERGENCY CLASSIFICATION LEVEL description for an ALERT.HA7: EAL-1 Selection Basis:N/AREFERENCE:N/A102 1 P age HS1ECL: SITE AREA EMERGENCYInitiating Condition: HOSTILE ACTION within tile PROTECTED AREA.Operating Mode Applicability: ALLEmergency Action Level:(1) A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reportedby ANY of the following personnel in Table HI:Table Hi: Security Supervision* Security Force Supervisor* Acting Security Manager* Security ManagerBasis:This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This eventwill 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 isessential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for theSecuritv Plan, Training and Qualification Plan, Sqfeguards Contingency Plan [and INDEPENDENTSPENT FUEL STORAGE INSTALL4 TION Security Program].As time and conditions allow, these events require a heightened state of readiness by the plant staff andimplementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The SITE AREAEMERGENCY declaration will mobilize ORO resources and have them available to develop andimplement public protective actions in the unlikely event that the attack is successful in impairingmultiple safety functions.This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise arenot a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a smallaircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of eventsis 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 incorporateSecurity-sensitive information. This includes information that may be advantageous to a potentialadversary, such as the particulars concerning a specific threat or threat location. Security-sensitiveinfonnation 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 wereincluded since any of these positions could be activated prior to meeting this EAL. The Security ForceSupervisor is a 24-hour position, the Acting Security Manager is activated after an Unusual Event hasbeen declared and the Security Manager is activated after an Alert is declared.REFERENCES:I. OERPOI-ZV-SH03, Rev. 12, Acting Security Manager2. OERPO1-ZV-TS08, Rev. 16, Security Manager104 P ag, e HS6ECL: SITE AREA EMERGENCYInitiating Condition: Inability to control a key safety function from outside the Control Room.Operating Mode Applicability: ALLEmergency Action Level:Note: The Emergency Director should declare the SITE AREA EMERGENCY promptly upondetermining 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 AuxiliaryShutdown Panel (ASP).ANDb. Control of ANY of the following key safety functions in Table H2 is not reestablished within 15minutes in Modes 1, 2 or 3 ONLY.Table H2: Safety Functions* Reactivity control* Core cooling* RCS heat removalBasis:This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternatelocations, and the control of a key safety function cannot be reestablished in a timely manmer. The failureto gain control of a key safety function following a transfer of plant control to alternate locations is aprecursor 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 onEmergency Director judgment. The Emergency Director is expected to make a reasonable, informedjudgment within 15 minutes whether or not the operating staff has control of key safety functions fromthe 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, asthe location where plant control is transferred in the event of a Control Room evacuation. The 15 minutetimeframe to control the key safety functions is identified as site specific information. The modeapplicability conditioning statement for Table H2 is based on the Technical Specification Operabilityrequirement 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 Evacuation2. Technical Specification 3.3.3.5 Remote Shutdown System106 1 P a o e HS7ECL: SITE AREA EMERGENCYInitiating Condition: Other conditions exist which in the judgment of the Emergency Director warrantdeclaration of a SITE AREA EMERGENCY.Operating Mode Applicability: ALLEmergency Action Level:(1) Other conditions exist which in the judgment of the Emergency Director indicate that events arein progress or have occurred which involve actual or likely major failures of plant functionsneeded for protection of the public or HOSTILE ACTION that results in intentional damage ormalicious 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. ANYreleases are not expected to result in exposure levels which exceed EPA PROTECTIVE ACTIONGUIDELINE exposure levels beyond the SITE BOUNDARY.Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declarationof an emergency because conditions exist which are believed by the Emergency Director to fall under theEMERGENCY CLASSIFICATION LEVEL description for a SITE AREA EMERGENCY.HS7: EAL-1 Selection Basis:N/AREFERENCE:N/A107 1 P a -e HG1ECL: GENERAL EMERGENCYInitiating Condition: HOSTILE ACTION resulting in loss of physical control of the FACILITY.Operating Mode Applicability: ALLEmergency Action Level:(1) a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reportedby ANY of the following in Table HI1:Table HI: Security Supervision* Security Force Supervisor* Acting Security Manager* Security ManagerANDb. EITHER of the following has occurred:1. ANY of the following safety functions in Table H2 cannot be controlled or maintained inMODES 1,2 or 3 ONLY.Table H2: Safety Functions* Reactivity control" Core cooling* RCS heat removalOR2. 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 tothe 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 orIMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system (e.g., pumps, heatexchangers., controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot bemaintained.Timely and accurate communications between Security Shift Supervision and the Control Room isessential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for theSecuritv Plan, Training and Qualification Plan, Safeguards Contingency Plan [and INDEPENDENTSPENT FUEL STOR4GE INSTALLA TION Securiz, Program].108 1 P a o e Emergency plans and implementing procedures are public documents; therefore, EALs do not incorporateSecurity-sensitive information. This includes information that may be advantageous to a potentialadversary, such as the particulars concerning a specific threat or threat location. Security-sensitiveinformation is contained in the Security Plan.HG1: EAL-1 Selection Basis:The positions of Security Force Supervisor, Acting Security Manager, and Security Manager were alsoincluded since any of these positions could be activated prior to meeting this EAL. The modeapplicability conditioning statement for Table H2 is based on the Technical Specification Operabilityrequirement 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 Manager2. OERPOI-ZV-TS08, Rev. 16, Security Manager3. Technical Specification 3.3.3.5 Remote Shutdown System109 P a -e HG7ECL: GENERAL EMERGENCYInitiating Condition: Other conditions exist which in the judgment of the Emergency Director warrantdeclaration of a GENERAL EMERGENCY.Operating Mode Applicability: ALLEmergency Action Level:(1) Other conditions exist which in the judgment of the Emergency Director indicate that events arein progress or have occurred which involve actual or IMMINENT substantial core degradation ormelting with potential for loss of containment integrity or HOSTILE ACTION that results in anactual loss of physical control of the FACILITY. Releases call be reasonably expected to exceedEPA PROTECTIVE ACTION GUIDELINE exposure levels offsite for more than the immediatesite area.Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declarationof an emergency because conditions exist which are believed by the Emergency Director to fall tinder theEMERGENCY CLASSIFICATION LEVEL description for a GENERAL EMERGENCY.HG7: EAL-1 Selection Basis:N/AREFERENCE:N/A110 1pa e 11 SYSTEM MALFUNCTION ICS/EALSTable S-1: Recognition Category "S" Initiating Condition MatrixUNUSUAL EVENTALERTSITE AREAEMERGENCYSU1 Loss of ALL offsiteAC power capability toemergency buses for 15minutes or longer.Op. Modes: 1,2,3,4SU2 UNPLANNED lossof Control Roomindications for 15 minutesor longer.Op. Modes: 1,2,3,4SU3 Reactor coolantactivity greater thanTechnical Specificationallowable limits.Op. Modes: 1,2,3,4SU4 RCS leakage for 15minutes or longer.Op. Modes: 1,2,3,4SU5 Automatic or manualtrip fails to shutdown thereactor.Op. Modes: 1,2SA1 Loss of ALL but one ACpower source to emergencybuses for 15 minutes or longer.Op. Modes: 1,2,3,4SA2 UNPLANNED loss ofControl Room indications for15 minutes or longer with asignificant transient inprogress.Op. Modes: 1,2,3.4SS1 Loss of ALL offsiteand ALL onsite ACpower to emergencybuses for 15 minutes orlonger.Op. Modes: 1,2,3,4GENERAL EMERGENCYSG1 Prolonged loss of ALLoffsite and ALL onsite ACpower to emergency buses.Op. Modes.: 1,2,3,4SA5 Automatic or manual tripfails to shutdown the reactor,and subsequent manual actionstaken at the reactor controlpanels are not successful inshutting down the reactor.Op. Alodes: 1,2SS5 Inability toshutdown the reactorcausing a challenge tocore cooling or RCSheat removal.Op. Af-odes: 1,2III IP a" e Table S-1: Recognition Category "S" Initiating Condition Matrix (cont.)UNUSUAL EVENT ALERT SITE AREA GENERALEMERGENCY EMERGENCYSU6 Loss of ALL onsite oroffsite communicationscapabilities.Op. Modes: 1,2,3.4SU7 Failure to isolatecontainment or loss ofcontainment pressurecontrol. 1,2,3,4SS8 Loss of ALL Vital DC SG8 Loss of ALL AC andpower for 15 minutes or Vital DC power sources forlonger. 15 minutes or longer. Op.Op. Modes: 1,2,3,4 MIodes: 1,2,3,4SA9 Hazardous eventaffecting a SAFETYSYSTEM needed for thecurrent operating mode.Op. Modes: 1,2,3,4112 P age SulECL: UNUSUAL EVENTInitiating Condition: Loss of ALL offsite AC power capability to emergency buses for 15 minutes orlonger.Operating Mode Applicability: 1, 2, 3, 4Emergency Action Level:Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that15 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 orlonger.Basis:This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plantmore vulnerable to a complete loss of power to AC emergency buses. This condition represents apotential reduction in the level of safety of the plant.For emergency classification purposes, "capability" means that an offsite AC power source(s) is availableto 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/AREFERENCES:I. OPOP04-AE-0001, Rev. 44, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus2. OPOP04-AE-0004, Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus3. 0PSP03-EA-0002, Rev. 32, ESF Power Availability4. Drawing 00000EOAAAA, Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. 1&2113 1 P a o e SU2ECL: UNUSUAL EVENTInitiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer.Operating Mode Applicability: 1, 2, 3, 4Emergency Action Level:Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that15 minutes has been exceeded, or will likely be exceeded.(1) An UNPLANNED event results in the inability to monitor one or more of the followingparameters 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 steamgenerators* Steam Generator Auxiliary FeedWater FlowBasis:This IC addresses the difficulty associated with monitoring normal plant conditions without the ability toobtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to amore 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 parameterscannot be determined from within the Control Room. This situation would require a loss of all of theControl Room sources for the given parameter(s). For example, the reactor power level cannot bedetermined 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 inaccordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC eventreport is required. The event would be reported if it significantlyimpaired the capability to performemergency assessments. In particular, emergency assessments necessary to implement abnormaloperating procedures, emergency operating procedures, and emergency plan implementing proceduresaddressing 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 ofreactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more ofthese parameters from within the Control Room is considered to be more significant than simply areportable 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 amain control board, the SPDS or the plant computer, tile availability of other parameter values may becompromised 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 steamgenerators is a site-specific parameter for the minimum number of steam generators needed for plantcooldown and shutdown.REFERENCES:1. OPOP05-EO-E020, Rev. 11, Faulted Steam Generator Isolation2. OPOP05-EO-FRHI, Rev. 23, Response to Loss of Secondary Heat Sink115 1 P a g e SU3ECL: UNUSUAL EVENTInitiating Condition: Reactor coolant activity greater than Technical Specification allowable limits.Operating Mode Applicability: 1,2, 3, 4Emergency Action Levels: (1 or 2)(1) RT-8039 reading greater than 30 pCi/cm3.(2) Sample analysis indicates that a reactor coolant activity value is greater than an allowable limitspecified in Technical Specifications.* Greater than I laCi/gm Dose Equivalent 1-131* Greater than 100/ E bar [iCi /gn gross activityBasis:This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in TechnicalSpecifications. This condition is a precursor to a more significant event and represents a potentialdegradation of the level of safety of the plant.Escalation of the EMERGENCY CLASSIFICATION LEVEL would be via ICs FAI or the RecognitionCategory R ICs.SU3: EAL-1 Selection Basis:RT-8039 is the Failed Fuel radiation monitor and samples via the CVCS letdown line. The value 30pCi/cm3 is the reading that is equivalent to I pCi/gm Dose Equivalent 1-131. The monitor value in thisEAL is the calculated monitor response if the RCS activity were equivalent to 1 [Ci/gm Dose Equivalent1-131. The value is based on Calculation STPNOC013-CALC-003. The value used in this EAL wasconservatively 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 orgreater than I 00/F bar puCi /gn gross activity.REFERENCES:1. Calculation No. STPNOC013-CALC-003 Rev.1, Gross Failed Fuel Monitor Response to RiseRCS Activity (RT-8039 EAL Threshold)2. STP Technical Specification Section 3/4.4.8 Specific Activity.116 1P a ge SU4ECL: UNUSUAL EVENTInitiating Condition: RCS leakage for 15 minutes or longer.Operating Mode Applicability: 1, 2, 3, 4Emergency Action Levels: (1 or 2 or 3)Note: The Emergency Director should declare the UNUSUAL EVENT promptly upon determining that15 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 orlonger.Basis:This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCSleakage has been detected and operators, following applicable procedures, have been unable to promptlyisolate the leak. This condition is considered to be a potential degradation of the level of safety of theplant.EAL #1 and EAL #2 are focused on a loss of mass from the RCS due to "unidentified leakage", "pressureboundary leakage" or "identified leakage" (as these leakage types are defined in the plant TechnicalSpecifications). EAL #3 addresses a RCS mass loss caused by an UNISOLABLE leak through aninterfacing 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 ControlRoom indications. Lesser values typically require time-consuming calculations to determine (e.g., a massbalance calculation). EAL #1 uses a lower value that reflects the greater significance of unidentified orpressure boundary leakage.The release of mass from the RCS due to the as-designed/expected operation of a relief valve does notwarrant an emergency classification. An emergency classification would be required if a mass loss iscaused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open andthe 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 CategoryRorF.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. 6requirements 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 throughany one steam generator, but the specification does not specify the type of leakage. Therefore, STPEGSwill use the leakage outside containment; which may include SG Tube Leakage, at 25 gpm for 15 minutesor 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 SU5ECL: UNUSUAL EVENTInitiating Condition: Automatic or manual trip fails to shutdown the reactor.Operating Mode Applicability: 1, 2Emergency Action Levels: (I or 2)Note: A manual action is ANY operator action, or set of actions, which causes the control rods to berapidly inserted into the core, and does not include manually driving in control rods or implementation ofboron injection strategies.(1) a. An automatic trip did not shutdown the reactor.ANDb. A subsequent manual action taken at the reactor control panels is successful in shutting down thereactor.(2) a. A manual trip did not shutdown the reactor.ANDb. EITHER of the following:I. A subsequent manual action taken at the reactor control panels is successful in shuttingdown the reactor.OR2. 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 thatresults in a reactor shutdown, and either a subsequent operator manual action taken at the reactor controlpanels or an automatic trip is successful in shutting down the reactor. This event is a precursor to a moresignificant 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 thereactor control panels to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actionsare successful in shutting down the reactor, core heat generation will quickly fall to a level within thecapabilities of the plant's decay heat removal systems.If an initial manual reactor trip is unsuccessfiul, operators will promptly take manual action at anotherlocation(s) on the reactor control panels to shut down the reactor (e.g., initiate a manual reactor trip) usinga different switch). Depending upon several factors, the initial or subsequent effort to manually trip thereactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a119 1 Pag e subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation willquickly 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 thecontrol rods to be rapidly inserted into the core (e.g., initiating a manual trip). This action does not includemanually 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 notconsidered to be "at the reactor control panels".The plant response to the failure of an automatic or manual reactor trip will vary based upon severalfactors including the reactor power level prior to the event, availability of the condenser, performance ofmitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manualactions taken at the reactor control panels are also unsuccessful in shutting down the reactor, then theEMERGENCY CLASSIFICATION LEVEL will escalate to an ALERT via IC SA5. Depending upon theplant response, escalation is also possible via IC FA 1. Absent the plant conditions needed to meet eitherIC 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), thefollowing classification guidance should be applied.* If the signal causes a plant transient that should have included an automatic trip and the RPS fails toautomatically shutdown the reactor, then this IC and the EALs are applicable, and should beevaluated.* 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 iswarranted.SU5: EAL-1 EAL-2 Selection Basis:N/AREFERENCES:I. OPOP03-ZG-0004, Rev. 45, Reactor Startup2. 0POP03-ZG-0005, Rev. 86, Plant Startup to 100%120 1 Page SU6ECL: UNUSUAL EVENTInitiating Condition: Loss of ALL onsite or offsite communications capabilities.Operating Mode Applicability: 1, 2, 3, 4Emergency 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 methodslisted in Table S2.(3) Loss of ALL of the following NRC communications methods listed in Table S2.Table S2: Communications MethodsEAL-1 EAL-2 EAL-3ONSITE 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 XCity" Microwave Lines to Houston X X* Security radio to Matagorda County X* Dedicated Ring-down lines X* ENS line XBasis:This IC addresses a significant loss of on-site or offsite communications capabilities. While not a directchallenge 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 communicationspossible (e.g., use of non-plant, privately owned equipment, relaying of on-site information viaindividuals 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 plantoperations.121 IPaoe EAL #2- addresses a total loss of the communications methods used to notify all OROs of an emergencydeclaration. The OROs referred to here are Matagorda County Sheriff s Office, and Texas Department ofPublic Safety Disaster District in Pierce.EAL #3- addresses a total loss of the communications methods used to notify the NRC of an emergencydeclaration.SU6: EAL-1, EAL-2, EAL-3 Selection Basis:Lines not included for offsite communications to ORO and NRC included links that would need relayingof information. Links were obtained from procedures OPGP05-ZV-0011, Emergency Communications.REFERENCES:1. OPGP05-ZV-001 1, Emergency Communications122 1 P a g e SU7ECL: UNUSUAL EVENTInitiating Condition: Failure to isolate containment or loss of containment pressure control.Operating Mode Applicability: 1, 2, 3, 4Emergency Action Levels: (1 or 2)(1) a. Failure of containment to isolate when required by an actuation signal.ANDb. ALL required penetrations are not isolated within 15 minutes of the actuation signal.(2) a. Containment pressure greater than 9.5 psig.ANDb. 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 whenrequired by an actuation signal. It also addresses an event that results in high containment pressure with aconcurrent failure of containment pressure control systems. Absent challenges to another fission productbarrier, 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/accidentcondition (e.g., a safety injection or high containment pressure); a failure resulting from testing ormaintenance 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 plantAOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate therequired penetrations, if possible.EAL #2- addresses a condition where containment pressure is greater than the setpoint at whichcontainment energy (heat) removal systems are designed to automatically actuate, and less than one fulltrain of equipment is capable of operating per design. The 15-1minute criterion is included to allowoperators time to manually start equipment that may not have automatically started, if possible. Theinability 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 aconcurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.1231 Paoe SU7: EAL-1 Selection Basis:N/ASU7: EAL-2 Selection Basis:If containment pressure reaches 9.5 psig, Containment Spray will actuate. If no train of ContainmentSpray is operating per design, the ability to lower containment pressure is compromised. One train ofContainment Spray (Technical Specifications 3/4.6.2) is defined as one containment spray systemcapable 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 Tree2. OPOP05-EO-FRZ1, Rev. 9, Response to High Containment Pressure3. Technical Specifications 3/4.6.21241 P a o e SA1ECL: ALERTInitiating Condition: Loss of ALL but one AC power source to emergency buses for 15 minutes orlonger.Operating Mode Applicability: 1, 2. 3, 4Emergency Action Level:Note: The Emergency Director should declare the ALERT promptly upon determining that 15 minuteshas 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 for15 minutes or longer.ANDb. ANY additional single power source failure will result in a loss of ALL AC power to SAFETYSYSTEMS.Basis:This IC describes a significant degradation of offsite and onsite AC power sources such that anyadditional 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. ThisIC provides an escalation path from IC SU 1.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying requiredpower 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., anonsite 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 emergencybuses 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 Bus2. OPOP04-AE-0004, Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus3. OPSP03-EA-0002, Rev. 32, ESF Power Availability4. Drawing 00000EOAAAA, Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. I126 Pa e c SA2ECL: ALERTInitiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer witha significant transient in progress.Operating Mode Applicability: 1, 2, 3, 4Emergency Action Level:Note: The Emergency Director should declare the ALERT promptly upon determining that 15 minuteshas been exceeded, or will likely be exceeded.(1) a. An UNPLANNED event results in tile inability to monitor one or more of the followingparameters 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 steamgenerators* Steam Generator Auxiliary FeedWater FlowANDb. 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) actuationBasis:This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during atransient 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 thusrepresents 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 parameterscannot be determined from within the Control Room. This situation would require a loss of all of theControl Room sources for the given parameter(s). For example, the reactor power level cannot bedetermined 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 inaccordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC eventreport is required. The event would be reported if it significantly impaired the capability to performemergency assessments. In particular, emergency assessments necessary to implement abnormaloperating procedures, emergency operating procedures, and emergency plan implementing proceduresaddressing 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 ofreactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more ofthese parameters from within the Control Room is considered to be more significant than simply areportable 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 amain control board, the SPDS or the plant computer, the availability of other parameter values may becompromised 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 andshutdown.REFERENCES:I. OPOP05-EO-EO20, Rev. 11, Faulted Steam Generator Isolation2. OPOP05-EO-FRHI1, Rev. 23, Response to Loss of Secondary Heat Sink1281 P a ge SA5ECL: ALERTInitiating Condition: Automatic or manual trip fails to shutdown the reactor, and subsequent manualactions taken at the reactor control panels are not successful in shutting down the reactor.Operating Mode Applicability: 1, 2Emergency Action Level:Note: A manual action is ANY operator action, or set of actions, which causes the control rods to berapidly inserted into the core, and does not include manually driving in control rods or implementation ofboron injection strategies.(1) a. An automatic or manual trip did not shutdown the reactor.ANDb. 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 thatresults in a reactor shutdown, and subsequent operator manual actions taken at the reactor control panelsto shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantialdegradation of the level of safety of the plant. An emergency declaration is required even if the reactor issubsequently shutdown by an action taken away from the reactor control panels since this event entails asignificant failure of the RPS.A manual action at the reactor control panels is any operator action, or set of actions, which causes thecontrol rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does notinclude 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 thereactor control panels (e.g., locally opening breakers). Actions taken at back-panels or other locationswithin the Control Room, or any location outside the Control Room, are not considered to be "at thereactor control panels".The plant response to the failure of an automatic or manual reactor trip will vary based upon severalfactors including the reactor power level prior to the event, availability of the condenser, performance ofmitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown thereactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safetyfunctions, the EMERGENCY CLASSIFICATION LEVEL will escalate to a SITE AREA EMERGENCYvia 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 accordancewith the Recognition Category F ICs; however, this IC and EAL are included to ensure a timelyemergency declaration.129 1 P age A reactor shutdown is detenmined in accordance with applicable Emergency Operating Procedure criteria.SA5: EAL-1 Selection Basis:N/AREFERENCES:1. OPOP05-EO-FRSl, Rev. 17, Response to Nuclear Power Generation -ATWS130 Paoe SA9ECL: ALERTInitiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the currentoperating mode.Operating Mode Applicability: 1, 2, 3, 4Emergency 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 ManagerANDb. EITHER of the following:1. Event damage has caused indications of degraded performance in at least one train of aSAFETY SYSTEM needed for the current operating mode.OR2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component orstructure needed for the current operating mode.Basis:This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structurecontaining SAFETY SYSTEM components, needed for the current operating mode. This conditionsignificantly reduces the margin to a loss or potential loss of a fission product barrier, and thereforerepresents 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 indicationsfor it will be readily available. The indications of degraded performance should be significant enough tocause 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 orreadily 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 reportinformation. This is intended to be a brief assessment not requiring lengthy analysis or quantification ofthe 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 breachalong the north wall which was included because it is a credible hazard and analyzed in the STPEGSUFSAR.REFERENCES:1. STPEGS UFSAR, Section 3.4.1, Flood Protection132 1P a g e SS1ECL: SITE AREA EMERGENCYInitiating Condition: Loss of ALL offsite and ALL onsite AC power to emergency buses for 15minutes or longer.Operating Mode Applicability: 1, 2, 3. 4Emergency Action Level:Note: The Emergency Director should declare the SITE AREA EMERGENCY promptly upondetermining 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 15minutes or longer.Basis:This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMSrequiring electric power including those necessary for emergency core cooling, containment heatremoval/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission productbarrier monitoring capabilities may be degraded under these conditions.This IC represents a condition that involves actual or likely major failures of plant functions needed forthe 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/AREFERENCES:I. OPOP04-AE-0001, Rev. 44, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus2. OPOP04-AE-0004, Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus3. OPSP03-EA-0002, Rev. 32, ESF Power Availability4. Drawing OOOOOEOAAAA, Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. 1&2133 P1 a o e SS5ECL: SITE AREA EMERGENCYInitiating Condition: Inability to shutdown the reactor causing a challenge to core cooling or RCSheat removal.Operating Mode Applicability: 1, 2Emergency Action Level:(1) a. An automatic or manual trip did not shutdown the reactor.ANDb. ALL manual actions to shutdown the reactor have been unsuccessful.ANDc. EITHER of the following conditions exists:" Core Cooling -Red entry conditions metOR* Heat Sink- Red entry conditions metBasis:This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip thatresults in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor areunsuccessful, and continued power generation is challenging the capability to adequately remove heatfrom the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions areunsuccessful 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 thatresulting firom an assessment of the plant responses and symptoms against the Recognition Category FICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additionalthreat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timelydeclaration 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 inabilityto 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 toremove heat from the RCS.134 1 P a g e
REFERENCES:
- 1. Procedure OPOP05-EO-F002, Rev. 2. Core Cooling Critical Safety Function Status Tree2. Procedure OPOP05-EO-F003, Rev. 6, Heat Sink Critical Safety Function Status Tree1351P age SS8ECL: SITE AREA EMERGENCYInitiating Condition: Loss of ALL Vital DC power for 15 minutes or longer.Operating Mode Applicability: 1, 2, 3, 4Emergency Action Level:Note: The Emergency Director should declare the SITE AREA EMERGENCY promptly upondetermining 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 15minutes or longer.Basis:This IC addresses a loss of Vital DC power which compromises the ability to monitor and controlSAFETY SYSTEMS. In miodes above Cold Shutdown, this condition involves a major failure of plantfunctions 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.3and determined to be 105.5 volts. At 105.5 volts or less, OPOP05-EO-ECOO, Loss of All AC Powerdirects the operators to open the battery output breakers.REFERENCES:1. OPOP05-EO-ECOO, Rev. 23, Loss of All AC Power1361 IP a ( e SG1ECL: GENERAL EMERGENCYInitiating Condition: Prolonged loss of ALL offsite and ALL onsite AC power to emergency buses.Operating Mode Applicability: 1, 2, 3, 4Emergency Action Level:Note: The Emergency Director should declare the GENERAL EMERGENCY promptly upondetermining 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.ANDb. 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 metBasis:This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC powercompromises the performance of all SAFETY SYSTEMS requiring electric power including thosenecessary for emergency core cooling, containment heat removal/pressure control, spent fuel heatremoval and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or morefission product barriers. In addition, fission product barrier monitoring capabilities may be degradedunder these conditions.The EAL should require declaration of a GENERAL EMERGENCY prior to meeting the thresholds forIC 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 projectedthat power cannot be restored to at least one AC emergency bus by the end of four (4) hours. Beyond thistime, plant responses and event trajectory are subject to greater uncertainty, and there is a higherlikelihood of challenges to multiple fission product barriers.The estimate for restoring at least one emergency bus should be based on a realistic appraisal of thesituation. Mitigation actions with a low probability of success should not be used as a basis for delaying aclassification 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 inparameters 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 entryconditions (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 throughinverters, or by Alternate AC (AAC) sources as defined in NUMARC 87-00. The STPEGS StationBlackout position credits any one of the three Standby Diesel Generators as the AAC source. Therequired 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 thebasis 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 Bus2. OPOP04-AE-0004, Rev. 15, Loss of Power to One of More 4.16 KV ESF Buses3. OPSP03-EA-0002, Rev. 32, ESF Power Availability4. Drawing OOOOOEOAAAA, Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. 1&25. OPOP05-EO-F002., Rev. 2, Core Cooling Critical Safety Function Status Tree6. OPOP05-EO-ECOO, Rev. 23, Loss of All AC Power7. STPEGS UFSAR Section 8.3.4, Station Blackout138 [P a a e SG8ECL: GENERAL EMERGENCYInitiating Condition: Loss of ALL AC and Vital DC power sources for 15 minutes or longer.Operating Mode Applicability: 1, 2, 3, 4Emergency Action Levels:Note: The Emergency Director should declare the GENERAL EMERGENCY promptly upondetermining 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 15minutes or longer.ANDb. Indicated voltage is less than 105.5 VDC on ALL Class 1E 125 VDC battery buses for 15minutes or longer.Basis:This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all ACpower compromises the performance of all SAFETY SYSTEMS requiring electric power including thosenecessary for emergency core cooling, containment heat removal/pressure control, spent fiuel heatremoval and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor andcontrol SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challengesto 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 atFukushima Daiichi. Minimum voltage for Class 1E 125 VDC battery buses was determined in calculation13-DJ-006 Rev.3 and determined to be 105.5 volts. At 105.5 volts or less, 0POP05-EO-EC0O, Loss of AllAC 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 Bus2. OPOP04-AE-0004, Rev. 15, Loss of Power to One of More 4.16 KV ESF Buses3. OPSP03-EA-0002, Rev. 32, ESF Power Availability4. OPOP05-EO-EC0O, Rev. 23, Loss of All AC Power5. Drawing OOOO0EOAAAA, Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. 1&2139 [ P a -e APPENDIX A -ACRONYMS AND ABBREVIATIONSAC ................................................................................................................................. Alternating CurrentAOP ........................................................................................................... Abnormal Operating ProcedureATW S .............................................................................................. Anticipated Transient W ithout ScramCDE ................................................................................................................ Com m itted Dose EquivalentCFR ................................................................................................................ Code of Federal RegulationsCTM T/CNM T ......................................................................................................................... Containm entCSF ........................................................................................................................ Critical Safety FunctionCSFST ................................................................................................. Critical Safety Function Status TreeDBA ......................................................................................................................... Design Basis AccidentDC ......................................................................................................................................... Direct CurrentEAL ...................................................................................................................... Emergency Action LevelECCS ...................................................................................................... Em ergency Core Cooling SystemECL ........................................................................................................... Emergency Classification LevelEOF ............................................................................................................ Emergency Operations FacilityEOP .......................................................................................................... Emergency Operating ProcedureEPA ........................................................................................................ Environmental Protection AgencyEPG ......................................................................................................... Emergency Procedure GuidelineERG .......................................................................................................... Emergency Response GuidelineFEM A ........................................................................................ Federal Emergency M anagem ent AgencyFSAR ............................................................................................................. Final Safety Analysis ReportGE .................................................................................................................... GENERA L EM ERGENCYIC ................................................................................................................................. Initiating ConditionID ....................................................................................................................................... Inside D iam eterISFSI ...................................................................................... Independent Spent Fuel Storage InstallationKeff .............................................................................................. Effective Neutron M ultiplication FactorLCO ......................................................................................................... Lim iting Condition of OperationLOCA .................................................................................................................. Loss of Coolant AccidentM SIV .............................................................................................................. M ain Steam Isolation ValveM SL ................................................................................................................................. M ain Steam LinemR, toRem , m rem , mREM ....................................................................... m illi-Roentgen Equivalent M anM W .............................................................................................................................................. M egawattNEI ........................................................................................................................ N uclear Energy InstituteNPP ............................................................................................................................ N uclear Power PlantNRC ......................................................................................................... Nuclear Regulatory Com m issionNSSS ........................................................................................................... Nuclear Steam Supply SystemNORA D ............................................................................ North American Aerospace Defense Com mand(NO )UE ................................................................................................... (Notification Of) Unusual EventNUM ARC ............................................................................ N uclear M anagement and Resources CouncilOBE ................................................................................................................ Operating Basis EarthquakeOCA ....................................................................................................................... Owner Controlled AreaODCM .................................................................................................... Offsite Dose Calculation M anualORO ........................................................................................................... Off-site Response OrganizationPA ........................................................................................................................................ Protected AreaPAG ................................................................................................................ Protective Action GuidelinePRA/PSA,.......................... .Probabilistic Risk Assessment / Probabilistic Safety Assessment1401Page PW R ................................................................................................................... Pressurized W ater ReactorPSIG ............................................................................................................ Pounds per Square Inch GaugeR .................................................................................................................................................... RoentgenRCS ....................................................................................................................... Reactor Coolant SystemRein, rein, REM ................................................................................................. Roentgen Equivalent M anRPS ................................................................................................................... Reactor Protection SystemRPV ........................................................................................................................ Reactor Pressure VesselRVW L ............................................................................................................. Reactor Vessel W ater LevelSAR ......................................................................................................................... Safety Analysis ReportSCBA ................................................................................................. Self-Contained Breathing ApparatusSG ..................................................................................................................................... Steam GeneratorSI ........................................................................................................................................ Safety InjectionSPDS ....................................................................................................... Safety Parameter Display SystemTEDE ........................................................................................................ Total Effective Dose EquivalentTOAF ............................................................................................................................. Top of Active FuelTSC ..................................................................................................................... Technical Support CenterW OG ............................................................................................................. W estinghouse Owners Group1411 Pao e APPENDIX B -DEFINITIONSThe following definitions are taken from Title 10, Code of Federal Regulations, and related regulatoryguidance documents.ALERT: Events are in progress or have occurred which involve an actual or potential substantialdegradation of the level of safety of the plant or a security event that involves probable life threateningrisk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases areexpected 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 IMMINENTsubstantial core degradation or melting with potential for loss of containment integrity or HOSTILEACTION that results in an actual loss of physical control of the facility. Releases can be reasonablyexpected 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 degradationof the level of safety of the plant or indicate a security threat to facility protection has been initiated. Noreleases of radioactive material requiring offsite response or monitoring are expected unless furtherdegradation of safety systems occurs.SITE AREA EMERGENCY: Events are in progress or have occurred which involve actual or likelymajor failures of plant functions needed for protection of the public or HOSTILE ACTION that results inintentional damage or malicious acts; I) toward site personnel or equipment that could lead to the likelyfailure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Anyreleases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond thesite 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 anInitiating Condition that, when met or exceeded, places the plant in a given EMERGENCYCLASSIFICATION LEVEL.EMERGENCY CLASSIFICATION LEVEL (ECL): One of a set of names or titles established by the USNuclear 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. TheEMERGENCY 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 thresholdindicating 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 fourEMERGENCY CLASSIFICATION LEVELS by virtue of tile potential or actual effects or consequences.Selected terms used in INITIATING CONDITION and EMERGENCY ACTION LEVELEMERGENCY ACTION LEVEL statements are set in all capital letters (e.g., ALL CAPS). These wordsare defined terms that have specific meanings as used in this document. The definitions of these terms areprovided below.CONFINEMENT BOUNDARY: The barrier(s) between spent fuel and the environment once the spentfuel is processed for dry storage.CONTAINMENT CLOSURE: Those actions necessary to place the RCB in the closedcontainment condition that provides at least one integral barrier to the release of radioactivematerial. Sufficient separation of the containment atmosphere from tile outside environment is tobe provided such that a barrier to the escape of radioactive material is reasonably expected toremain 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 threatmessage exists that supports the threat, or (3) A specific known group or organization claimsresponsibility 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 anelectrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically beconsidered an explosion. Such events may require a post-event inspection to determine if the attributes ofan 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 ofsufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to becomecompletely depressurizedFIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts oroverheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOTrequired if large quantities of smoke and heat are observed.HATCH MONITOR: Temporary monitor installed when Containment High Range Radiation MonitorsRT-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 thestation.143 11P a -e HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force todestroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includesattack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used todeliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTIONshould not be construed to include acts of civil disobedience or felonious acts that are not part of aconcerted 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 bystealth and deception, equipped with suitable weapons capable of killing, maiming, or causingdestruction.IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relativelyshort period of time regardless of mitigation or corrective actions.INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed andconstructed for the interim storage of spent nuclear fuel and other radioactive materials associated withspent fuel storage.NORMAL LEVELS: As applied to radiological IC/EALs, the highest reading in the past twenty-fourhours excluding the current peak value.OWNER CONTROLLED AREA: The area surrounding the PROTECTED AREA where STP NuclearOperating 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 protectiveactions to safeguard against radiation exposure from nuclear incidents.PROTECTED AREA: The area under continuous access monitoring and control, and armed protection asdescribed in the site Security Plan.REFUELING PATHWAY: Includes all the cavities, tubes, canals and pools through which inradiated fuelmay be moved, but not including the reactor vessel.RUPTURE(D): The condition of a steam generator in which primary-to-secondary leakage is of sufficientmagnitude to require a safety injection.SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing itin the cold shutdown condition., including the ECCS. These are typically systems classified as safety-relatedSECURITY CONDITION: Any Security Event as listed in the approved security contingency plan thatconstitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation tothe 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 securitysystem of reduction of the operational effectiveness of that system. A security event can result in either aSECURITY CONDITION or HOSTILE ACTION.SITE BOUNDARY: The edge of the plant property whose access may be controlled by STPEGS. Thisboundary 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) anexpected plant response to a transient. The cause of the parameter change or event may be known orunknown.THYROID CDE: The dose equivalent to the thyroid from an intake of radioactive material by anindividual during the 50-year period following the intake.VALID: An indication, report or condition is considered to be VALID when it is verified throughappropriate means such that there is no doubt regarding the indicator's operability, the condition'sexistence, 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 verificationmethods should be completed in a manner the supports timely emergency declaration.VISIBLE DAMAGE: Damage to a component or structure that is readily observable withoutmeasurements, testing, or analysis. The visual impact of the damage is sufficient to cause concernregarding the operability or reliability of the affected component or structure.145 1 P a g e
Attachment
3STPEGS Emergency Action Level Technical Bases Document -redline markup NEI 99-40 6]STPEGS Emergency Action LevelTechnical Bases Document Rev. 0hruI , QW.0 Raw' 61 ,mimnator~ !!-U+/- Key. b ImplementationI~-aa vn entnNovember 201 &T-TOMVJ1 Fet--7y 5NOTE: 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 YounigPreparatfico TeRamLa..m,' Bak..r .e.lon Nuclear./Co.pot..eC-raig B .anne. r 2 -PSG and Hope Cr-eek Nuclear Generating Stations/USAJohn Egdorf ...... an Power StationJack Lewis Entergy NuElear-'/CoEperateC. Kelly W.alk.. r Oprations Suppon. N6e.Rev~ew- eamChisBone Southom Nucleor/CorporoleBill Chal-sse Encrooen Sorv~icee, hie.KenHt Crooker-- PrFogeFss EnRoo'B9Frunewie- Nuclolar PlantDon Crow.l D-ue Energy/C-orporateRogor Freontian Cons~tellto Enr- Nuolar Group/CorPorateW.A,7alt ... Lo A Nuclear/Co rporatoKen Meade FENOC/CorporateDon Mothena NctkrAFa Energy/Coporate-David StEbaugh! EP Consulting, LLC-a-ee --a--i- Diabloeano Powor Plant/STARSNOTICE-".'-", !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. Thisseherne is an funldamental componient ofan emergency plan inthat itprovides the eie thresqholds thlat vill alias% Site personnel to rapidlyfim~pl~emeiita angeeOf PH ,ie4i~ed eflwffgelizff pOnSP)e mleasur-es. Ani emerfgency classification scheme also theilitates tinmcl'; decisiei ntakine byan 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 asitespeeifi emfffergenc classification schemfe. The methaodaky described in this dOeumlrt iS Eeon itInt vAith Federal regulationls, and relatedi U-SNuclear Regulatory Commission (NRCIF~~ requireents and gui dance. In particular, this me~tho~dology has been! endorsed by the NRC as anlecepieble-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. oNPt 99 01 oonbinsm a set of g@-eneri InitifltiogL C-ondiOIlL (0), Effergency Action Lev ela (FAIs) and fissionl product barrier status thresholds. itAsO includes supporting technical basis information, developer notes and reEomm-e.nded classification instr-cti'Os for usFers. User.s ,shouldi1plement ... , 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..snecessan; 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 readilybSer.'able PA~ an Fd thresholds. OtherF benelf~its inlude the developtoent of a Sound basis docuenllit, thle adoptOian Of ilidulstfr stanldardinStruc.tions for1 .emerg.-ency.c 'lassificat in (e.g., tran.sient events. classification of.multiple ev ents, upgrf.ading., g rig.... e .incorpationof features to improve human performance. An emergency c.,..lassiOicatio using this schem .sill be appropriate to the risk pod to plant w.okersand 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 NRCrequirement.an. d guidance prior to beginning their effots. Questions concerning this document may he directed t0 the NE! E:ms rgencyPreparedness 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~orationiOf .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 requireare. 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 CONTENTSEXECUTIVE SUMMARY ....................................................................................................... 1I REGULATORY BACKGROUND ...................................................................1 DEVELOPMENT OF EMERGENCY ACTION LEVELS ........................................................ 11.1 O PE R A T IN G RE A C T O R S ................................................................................................................ 11.1 REG U LA TO R Y B A C K G RO UN D .................................................................................................... 11 D2 PERMANENTLY DEFUELED STATION i1.3 I3N) S.PENT FU .EL-T STOPA 6 .GE I NsTTA11 AT I ON (Is FSI) 11.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ..................................... 21.i N R C O R D ,ER E A 12 051 ................................................................................................................... 21.3 N RC O RD ER EA -12-051 ............................... ...................................... 31.5 APPLICABILITY TO ADVANCED ADD SMALL MODULAR EACTRS .............................. 12 KEY TERMINOLOGY USED IN NEI 99-01............... .. .................42 KEY TERM INOLOGY ......................................................................................................... 42.1 EMERGENCY CLASSIFICATION LEVEL (ECL) ....................................................................... 52.2 IN ITIATIN G CON D ITIO N (IC) .................................................................................................. 62.3 EM ERGENCY ACTION LEVEL (EAL) ........................................................................................ 62.4 FISSION PRODUCT BARRIER THRESHOLD ........................................................................... 73 DESIGN OF THE NEI 99-01 EMERGENCY CLASSIFICATION SCHEME ..............73 DESIGN OF THE STPEGS EMERGENCY CLASSIFICATION SCHEME ............................ 83.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS) ................................83.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS ..................... 113.3 NSSS DESIGN DIFFERENCES ........................................................................... i133.3 STPEGS DESIGN CON SIDERA TIONS ..................................................................................... 113.4 ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION ........................... 123.5 IC AND EAL MODE APPLICABILITY ..........................................134 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 ................................................................................... 154.2 CRITICA L CHA RA CTERISTICS ................................................................................................ 164.3 INSTRUM ENTATION USED FOR EALS ................................................................................. 174.4 PRESENTATION OF SCHEME INFORM.ATIO.. ..TO USER ....................................................... 20'1.5 INTEGRATION OF ICSiEALS WITH PLANT PROCEDURES...................................... 2141.6 BASIS DOCUMENT .............................................. .............4.7 DEVELOPER AND USER. FEED..AC. **--'*.... ....4.4 REFERENCES TO STPEGS AOPS AND EOPS-. ....................................................................... 195 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ........................................ 195.1 GENERAL CONSIDERATIONS ........................... ...................................................... 195.2 CLASSIFICATION M ETHODOLOGY ..................................................................................... 20 5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS ....................................... 205.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION ............................. 215.5 CLASSIFICATION OF IMMINENT CONDITIONS ". 1...........................215.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING ............ 225.7 CLASSIFICATION OF SHORT-LIVED EVENTS ............................ .......................................225.8 CLASSIFICATION OF TRANSIENT CONDITIONS ................................. ........... 225.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION ....... 235.10 RETRACTION OF AN EMERGENCY DECLARATION ...................................................... 236 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/EALS .......................... 247 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS ................. 54558 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS ............. 8999 FISSION PRODUCT BARRIER ICS/EALS ................................................... ............... 929310 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS .. 128I-,O11 SYSTEM MALFUNCTION ICS/EALS .................................................................... 1601-62APPENDIX A -ACRONYMS AND ABBREVIATIONS ................................................ 200202APPENDIX B -DEFINITIONS ................................................................................... 202204APPENDIX C -PERMANENTLY DEFUELED STATION ICs/EALs. 193 THIS PAGE IS LEFT INTENTIONALLY BLANK DEVELOMENT OF EMERGENCY ACTION LEVELSFOR NON-PASSIVE REACTORS1 REGULATORY BACKGROUND1 DEVELOPMENT OF EMERGENCY ACTION LEVELS1. 1 OPERATING PEACTORS1.1 REGULATORY BACKGROUNDTitle 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 anemergency classification scheme. A review of the relevant sections listed below will aid the reader inunderstanding 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 OrganizationThe above regulations are supplemented by various regulatory guidance documents. Three documents ofparticular relevance to NEI 99-01 are:* NUREG-0654/FEMA-REP-I, Criteria for Preparation and Evaluation of Radiological Emergency ResponsePlans 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 withlicensing/regulatory complianec per-sonnel to idcntiby and uinderstand all applic-able requtirementsanguidance. may als..... bhe ddi.racte.ed to the NEI Emergencv Prepar'edness staff.1.2 PERMANENTLY DEFUELED STA TIONNEI 99 01 proevides guidance for an emergIency classification seheme applicable to a permanently defuieledstation. This is a station that generated spent fuel under a 10 CFR § 50 license, has permaniently ceaseoperations and- w.ill stare the spent fuel onsite for an extended period of time. The emer-gency classific-ation levelsI Wpage applicable k) this ty.pe of StatiOn _ae cAontn Wi the., requirements of 10 DFR § 50 and the guidance inX1ITI Cf- E %C-CA IE'I, A A D UD Iin order te relax the emergency plan requiremants applicable to an operating statin, the owner of a permfanentlydefueled station must demn!Ostrate that no crdible event can result in a significant radiological r-elease beyondthe site boundary. it is expeted that this erSTPificSai will conwfirm that the source tedrea and motive fce availablein the perimanenly defeled ofndition aFre insufficient to warrant classifications of a Site Afiea Emegeney orGeneral 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 anlAler1 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 feetwest ofthe Unit 2 Reactor Building. The STP ISFSI will be within the site Protected Area and is scheduled to beoperational in 2016.Selected guidance in NEI 99-01 is applicable to the STPEGS licensees electing, to usel10 CER 50 emergency planto fulfill the requirements of 10 CFR 72.32 for a stand-alone ISFSI. The eSergency classification levelsapplicable to an ISFSI are consistent with the requirements of 10 CFR § 50 and the guidance in NUREG0654/FEMA-REP-l. The initiating conditions germane to a 10 CFR § 72.32 emergency plan (as described inNUREG-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 coversthe spectrum of credible natural and man-made events included within the scope ofoat -the STPEGS ISFSd Idesign.. This IC is not applicable to installations or fclitathat ma:d prcess andr repaclage spent fuel (e.g., amoaitmred Retrievable Stoaage Faeility orf an pSFSi at a spent ftel preeessine i addition, appropriateaspects of IC HUI and IC HAl sheuld also be ineluded to address a HOSTILE ACTION directed against a-theSTPEGS ISFSI.The analysis of potential onsite and offsite consequences of accidental releases associated with the operation ofan ISFSI is contained in NUREG-l 140, A RegilatoyjAnal'sis on Emnergenci.' Pr-epar-edness for- Futel C'ycle andOthe1 r Radioactive Al.atergial Licensees. NeREG-l 140 concluded that the postulated worst-case accidentinvolving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that themaximum offsite dose to a member of the public due to an accidental release of radioactive materials Would notexceed I rem Effective Dose Equivalent.Regarding the above information, the expectations for an offsite response to an Al-ert-ALERT classified under a10 CFR § 72.32 emergency plan are generally consistent with those for an-a Notification of Uusuqal Even.1tUNUSUAL EVENT in a 10 CFR § 50.47 emergency plan (e.g., to provide assistance if requested). Also, thelieensee's STPEGS Emergency Response Organization (ERO) required fora 10 CFR § 72.32 emergency plan isdifferent than that prescribed for a 10 CFR § 50.47 emergency plan (e.g., no emergency technical supportfunction).2 P a g e 1.4 NRT ORDER EA 12 051-1.3 NRC ORDER EA-12-051The Fukushimna Daiichi accident of March 11,2012, was the result of a tsunami that exceeded the plant's designbasis and flooded the site's emergency electrical power supplies and distribution systems. This caused anextended loss of power that severely compromised the key safety functions of core cooling and containmentintegrity, and ultimately led to core damage in three reactors. While the loss of power also impaired the spentfuel pool cooling function, sufficient water inventory was maintained in the pools to preclude fuel damage fromthe loss of cooling.Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessaryto ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR50.109(a)(4)(ii). Among them was to provide each spent fuel pool with reliable level instrumentation tosignificantly enhance the ability of key decision-makers to allocate resources effectively following a beyonddesign basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to Modi, Licenses withRegard to Reliable Spent Fuel Pool Instrunentation, on March 12, 2012, to all US nuclear plants with anoperating 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 inassociated spent fuel storage pools capable of supporting identification of the following pool water levelconditions by trained personnel: (1) level that is adequate to support operation of the normal fuel pool coolingsystem, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuelpool operating deck, and (3) level where fuel remains covered and actions to implement make-up water additionshould 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 theused 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 connectioncapability.NEI 12-02. Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regardto 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 spentfuel pool level instrumentation associated with NRC Order EA-12-051. These EALs are included within existingIC A-ARA2. and new ICsAS-2- RS2 and-AG- RG2. It is reco:mmr. ended that tThese EALs will be implementedwhen 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~echanges 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.e3[1P agRe 1.5 APPLICABILITY TO ADVANCE AND SMALL MAODULAR REACTODERCIQNThe 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 badapted to advan..ed non passive designs (often r.efcrred o as 3fg ....rati... I .l desig.s) as well. Developers oani emfer-gency classification scheme for- an advanced non passive reaetor- plant may need to proapose deviationsfroem the generici guidance to account for- the differences inq design par-ameters and criteria, and operatingcharacteristics and capabilities, between 2nJand-3rd generation plants.Thoce are significant design and operating difaferenes between lage commerial nuclear power plants (Mf anygenefration) and Small Modular- Reactor-s (Wk~s) (e.g., differences in source term). For this r-easoni,-4thisdocumient is not aeolieable to SMRs.ucy EIvnnpn nnnIULfl flE iagnE Eim mE~ CO Al-.p * * *u*um* B nww -.w *
- U
- u* wa2 KEY TERMINOLOGY USEDThere are several key terms that appear throughout the NEI 99 01 EAL methodology. These tenns are introducedin this section to support understanding of subsequent material. As an aid to the reader, the following table isprovided as an overview to illustrate the relationship of the terms to each other............. Level .EMERGENCY CLASSIFICATION LEVELAIei4ALERT SAE GE-veetUNUSUALEVENT+ 40Initiating Condition Initiating Condition Initiating Condition Initiating Condition+ .4 0Emergency Action Emergency Action Emergency Action Emergency ActionLevel (1) Level (1) Level (1) Level (1)" Operating Mode -Operating Mode
- Operating Mode
- Operating ModeApplicability Applicability Applicability Applicability" Notes
- Notes
- Notes
- Notes" Basis
- Basis
- Basis
- Basis(1) -When making an emergency classification, the Emergency Director must consider all infornationhaving a bearing on the proper assessment of an Initiating Condition. This includes the EmergencyAction Level (EAL) plus the associated Operating Mode Applicability, Notes and the infonning Basisinformation. In the Recognition Category F matrices, EALs are referred to as Fission Product BarrierThresholds; 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 onsiteand 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 plantor indicate a security threat to facility protection has been initiated. No releases of radioactive material requiringoffsite 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 eventinformation and decision-making.2.1.2 2.1.2Aeirt ALERTEvents are in progress or have occurred which involve an actual or potential substantial degradation of the levelof safety of the plant or a security event that involves probable life threatening risk to site personnel or damage tosite equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of theEPA PAG exposure levels.Purpose: The purpose of this classification is to assure that emergency personnel are readily available to respondif the situation becomes more serious or to perform confirmatory radiation monitoring if required, and provideoffsite 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 neededfor protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) towardsite 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 levelswhich exceed EPA PAG exposure levels beyond the site boundary.Purpose: The purpose of the Site Area En-ergenevSITE AREA deelarationEMERGENCY declaration is toassure that emergency response centers are staffed, to assure that monitoring teams are dispatched, to assure thatpersonnel 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 governmentauthorities.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 ormelting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss ofphysical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsitefor more than the immediate site area.Purpose: The purpose of the General Emergency GENERAL EMERGENCY declaration is to initiatepredetermined protective actions for the public, to provide continuous assessment of information from thelicensee and offsite organizational measurements, to initiate additional measures as indicated by actual orpotential releases, to provide consultation with offsite authorities, and to provide updates for the public throughgovernment authorities.--INITIATING CONDITION (IC)4-,32.2An 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.Discussion: An IC describes an event or condition, the severity or consequences of which meets the definition ofan emergency classification level. An IC can be expressed as a continuous, measurable parameter (e.g., RCSleakage), an event (e.g., an earthquake) or the status of one or more fission product barriers (e.g., loss of the RCSbarrier).Appendix I of NUREG-0654 does not contain example Emergency Action Levels (EALs) for each ECL, butrather Initiating Conditions (i.e., plant conditions that indicate that a radiological emergency, or events that couldlead to a radiological emergency, has occurred). NUREG-0654 states that the Initiating Conditions form the basisfor 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 anemer-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 statusindications; observable events; results of calculations and analyses; entry into particular procedures; and theoccurrence of natural phenomena.61Pa e 2.2.2 2-.4FISSION PRODUCT BARRIER THRESHOLDA pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission productbarrier.Discussion: Fission product barrier thresholds represent threats to the defense in depth design concept thatprecludes the release of radioactive fission products to the environment. This concept relies on multiple physicalbarriers., any one of which, if maintained intact, precludes the release of significant amounts of radioactivefission products to the environment. The primary fission product barriers are:* Fuel Clad* Reactor Coolant System (RCS)* ContainmentUpon determination that one or more fission product barrier thresholds have been exceeded., the combination ofbarrier loss and/or potential loss is compared to the fission product barrier IC/EAL criteria todetermine the appropriate ECL.In some accident sequences, the ICs and EALs presented in the Abnormal Radiation Levels/ RadiologicalEffluent (RA) Recognition Category will be exceeded at the same time, or shortly after, the loss of one or morefission product barriers. This redundancy is intentional as the former ICs address radioactivity releases that resultin certain offsite doses from whatever cause, including events that might not be fully encompassed by fissionproduct barriers (e.g., spent fuel pool accidents, design containment leakage following a LOCA, etc.).71Page 3 DESIGN OFTHE NEI 99-01 EMERGENCY CLASSIFICATION3 DESIGN OF THE STPEGS EMERGENCY CLASSIFICATION SCHEME-ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS)243.1An effective emergency classification scheme must incorporate a realistic and accurate assessment of risk, bothto plant workers and the public. There are obvious health and safety risks in underestimating the potential oractual 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 strikean appropriate balance between reasonably anticipated event or condition consequences, potential accidenttrajectories, and risk avoidance or minimization.There are a range of "non-emergency events" reported to the US Nuclear Regulatory Commission (NRC) staff inaccordance with the requirements of 10 CFR § 50.72. Guidance concerning these reporting requirements, andexample 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 theattributes of each ECL. The goal of this process is to answer the question, "What events or conditions should beplaced Linder each ECL?" The following sources provided information and context for the development of ECLattributes." 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 PlantsI lndustry Operating Experience* Input from ind'!Stf+' subject matter experts and NRC staff'membersat STPEGSThe following ECL attributes were created by the Revision 6 Preparation Team; to aid in the development of ICsand Emergency Action Levels (EALs). The team decided to include the attributes in this revision since they Theattributes may be useful in briefing and training settings (e.g., helping an Emergency Director understand why aparticular 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 limitedto 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 tileplant.(C) A consequence otherwise significant enough to warrant notification to local, State and Federalauthorities.2.1.23. 1.2 3..2 Alet ALERTAn A-ei4ALERT, as defined in section 2.1.2, includes but is not limited to an event or condition thatinvolves:(A) A loss or potential loss of either the fuel clad or Reactor Coolant System (RCS) fission productbarrier.(B) An event or condition that significantly reduces the margin to a loss or potential loss of the fuel clador RCS fission product barrier.(C) A significant loss of control of radioactive materials resulting in an inability to control radiationlevels within the plant, or a release of radioactive materials to the environment that could result indoses 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 !hcedifRccted 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 anevent 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 productbarriers within a relatively short period of time. Precursor events and conditions of this type includethose 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% ofan 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 toan 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 productbarriers. Precursor events and conditions of this type include those that lead directly to core damageand loss of containment integrity.(C) A release of radioactive materials to the environment that could result in doses greater than an EPAPAG at or beyond the site boundary.(D) A HOSTILE ACTION resulting in the loss of key safety functions (reactivity control, corecooling/RPV water level or RCS heat removal) or damage to spent fuel.3.1.5 4--4Risk-Informed InsightsEmergency preparedness is a defense-in-depth measure that is independent of the assessed risk from anyparticular accident sequence; however, the development of an effective emergency classification scheme canbenefit from a review of risk-based assessment results. To that end, the development and assignment of certainICs and EALs also considered insights from several site-specific probabilistic safety assessments (PSA -alsoknown 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 coredamage 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 HotShutdown, was assigned an ECL of Site Area Emergency. SITE AREA EMERGENCY. Precursorevents to a loss of all AC power were also included as an U:nSumal Event UNUSUAL EVENT and anA4e4-ALERT.A station blackout coping analyses performed in response to 10 CFR § 50.63 and Regulatory Guide1.155, Station Blackout, may be used to determine a time-based criterion to demarcaie between a 8i4eArea Emer'geneySlTE AREA EMERGENCY and a General EmergencyGENERAL EMERGENCY.The time dimension is critical to a properly anticipatory emergency declaration since the goal is tomaximize the time available for State and local officials to develop and implement offsite protectiveactions. 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 forthe 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 progressionsleading to containment failure. Because of these uncertainties, predicting the status of containmentintegrity may be difficult under severe accident conditions. This is why maintaining containmentintegrity alone following sequences leading to severe core damage is an insufficient basis for notescalating to a Gener-al EmergencyGENERAL EMERGENCY.3. PSAs indicated that leading contributors to latent fatalities were sequences involving a containmentbypass, a large Loss of Coolant Accident (LOCA) with early containment failure, a Station Blackoutlasting 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 timelyfashion.3.2 3-TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTIONLEVELSThe NEI 99 0! STPEGS methodology makes use of symptom-based, barrier-based and event-based ICs andEALs. Each type is discussed below.Symptomn-based ICs and EALs are parameters or conditions that are measurable over some range using plantinstrumentation (e.g., core temperature, reactor coolant level, radiological effluent, etc.). When one or more ofthese parameters or conditions are off-normal, reactor operators will implement procedures to identify theprobable cause(s) and take corrective action.Fission product barrier-based ICs and EALs are the subset of symptom-based EALs that refer specifically to thelevel of challenge to the principal barriers against the release of radioactive material from the reactor core to theenvironment. These barriers are the fuel cladding, the reactor coolant system pressure boundary, and thecontainment. 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 safetysignificance. These include the failure of an automatic reactor scram/trip to shut down the reactor, naturalphenomena (e.g., an earthquake), or man-made hazards such as a toxic gas release.3.3 NSSS DESIGN DIFFERENCE.SM--STPEGS DESIGN CONSIDERATIONSThe NE! 99 01 emcrgeney cl-assification Scemlffe accountS fcr the design difefcenees between PWIRs and BWA'sby speeifý'ing EALs uinique to each type of Nuclear Steam Supply System (NSSS). There are also significantdesign differencs aogPRNSs ther~efor-e, guidance is provided to aid in the development of EALsappropr~iate to different PAIR NSSS ty'pes. Where nlecessar~y, develcpment guidance also addresssuiuconsiderations 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 whenonverting, the ggeneric guidance of this documzent into a site specific classification scheme. The goalistmainitain as much fidelilty as possible to the intent of generici this end, developers of a schemie for _An -advancednon passive reactor- may need to add, modify, or delete some infen~atiin. contained in this document; thesechanges 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.AEmrecClassification Sc~heme for this ty'pe Of planlt shoul1d be developed in arceeoDcevelopmc;ew of E gcyieonLeve!s, Adwiacd Pas.qiivc Ligh Ut !facconstraints imposed by the plant design and ope.ating char'acteristics. Tov'41 ,,;1. 07 0"1, Alt.. ed^.1ieg....4"....PeocttefslCs anld EALs within the...... ......... ..I.. .IlIIP agRe 3.3The South Texas Project Electrical Generating Station (STPEGS) is composed of two units, each having anidentical 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 thecoolant. 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 heatgenerated in the core to the SGs using RCPs to circulate the water. RCS pressure is maintained by means of apressurizer attached to the hot leg of one of the loops. The RCS is designed to circulate borated demineralizedwater at temperatures. pressures and flow rates consistent with the design thermal and hydraulic performance ofthe NSSS.The Reactor Coolant Pressure Boundary Leak Detection System consists of temperature, level, humidity, andradioactivity sensors with associated instrumentation and alarms. Small leaks are detected by temperature andlevel changes of systems, increasing sump levels, and humidity and radioactivity concentration changes insidethe Containment. Large leaks are detected by changes in reactor coolant inventory, changes in flow rates inprocess lines and changes in sump level.Emergency Core Cooling System consists of three independent trains. each one capable of providing 100 percentof the required flow to the core in the unlikely event of a LOCA. Each train consists of one high-head safetyiniection pump and one low-head safety injection pump. Heat is removed from the system during recirculationby the residual heat removal heat exchangzer (low-head pump only). The piping and valving associated with eachof the three subsystems are identical. In the event of a steam pipe rupture. the ECCS provides adequate shutdowncapability.The Reactor Containment is a post-tensioned concrete cylinder with a steel liner plate, hemispherical top, andflat bottom. This structure provides a virtually leaktight barrier to prevent escape of fission products to theenvironment in the unlikely event of a loss of coolant accident (LOCA).--ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION3.4The 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 10S S -System Malfunction -Section II*PD Permaneiiflv Defuceled statben Atniefdix C12 P age Each Recognition Category section contains a matrix showing the ICs and their associated emeigeneyclassifiation lcvcls.EMERGENCY CLASSIFICATION LEVELS. The following information and guidance isprovided 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 tomeet the intent of the IC.Devclepefs should add... .a.. e.ample EAL. if the generi. appr.a.h to the development .an example EALcannot 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., potentialloss or loss). This presentation method shows the synergism among the thresholds, and supports accurateassessments.Basis -Provides background information that explains the intent and application of the IC and EALs. In somecases, the basis also includes relevant source information and references.incelude clarificaitions, r-efercenccs, e~amples, instructions for- calculations, etc. DeVeloper notes should notbe inechded On the site's emer-gency classification scheme basis doeumcnt. Develapers may clect to ineludcinfor-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 feorassi-anina- the WC to it given ECL.--IC AND EAL MODE APPLICABILITYQ-3.35The NEI 99 01 STPEGS emergency classification scheme was developed recognizing that the applicability ofICs and EALs will vary with plant mode. For example, some symptom-based ICs and EALs can be assessed onlyduring the power operations, startup, or hot standby/shutdown modes of operation when all fission productbarriers are in place, and plant instrumentation and safety systems are fully operational. In the cold shutdown andrefueling modes, different symptom-based ICs and EALs will come into play to reflect the opening of systemsfor routine maintenance, the unavailability of some safety system components and the use of alternateinstrumentation.The following table shows which Recognition Categories are applicable in each plant mode. The ICs and EALsfor a given Recognition Category are applicable in the indicated modes.13 Page MODE OF APPLICABILITY MATRIXRecognition Cate~orvMode AR C E F H P-1D SPower Operations X X X X XStartup X X X X XHot Standby X X X X XHot Shutdown X X X X XCold Shutdown X X X XRefueling X X X XDefueled X X X XD'l-fulleffly _ _ _ _-STPEGS Operating ModesMode Description Criteria (Rx Power excludes decay heat)1 Power Operations Reactor Power > 5%, Keff> 0.99 T Avi_> 350'F2 Startup Reactor Power < 5%, Keff> 0.99 T Avg > 3507FHot Standby Reactor Power 0% Keff< 0.99 T Avg > 350'FHot Shutdown Reactor Power 0% Keff< 0.99 3500F > T Avg > 200°FCold Shutdowvn Reactor Power 0% Keff< 0.99 T Avgy < 200°F6 Rfuelina Reactor Power 0% Keff <0.95 T Avg < 140'FFuel in the reactor vessel with the vessel head closure bolts less than fully tensionedor with the head removed.Defueled All fuel removed from the reactor vessel (i.e., full core offload during refuel orextended outage)Developers wiAl need to intorporate the mode .riteria fram unit speeific Tehinical Speeifieations into theiremerenc laifi n ..h.me. in addition, the sch.me must .also .icl.de the following miade deSs5pecific te N Hl 99 01:Deftieled (Noflre}7All~ fuel1 rprn'v'd ffa il,,o m-ei v1: -lr__ll I ý-I _ ._ý!.e --- -..ttktt tuitifttt ret~tlfix OF or dtrKiEWOoutage).14 Page
- 4. SITE-SPECIFIC SCHEME DEVELOPMENT GUIDANCE4 STPEGS SCHEME DEVELOPMENTeffiegefe n Iasification scheme. COHn ptuall)' the appi-each diScussed here mirror~s the approeach Elsed toprepare emetrgency' operzatinig procedures generic material prepared by reactor: vendor- owners goups Iscneritted by each nucelear: poer'e plant intot site specifie emier-gency cperatiing proeedures. Likewise, theemferg-encyý 6iassination Snemcl OeVeioper' WHI USe tile genieri. gUimoanee 1n NEI 99W tO prepare a site speecmclemerg-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 incompleteemerg ..ency classification scheme that will likely not .eceive the necessary reg.lat.y appreval-a.1 GENERALIMPLEMENTATION GU 1DANCE4.1 GENERAL DEVELOPMENT PROCESSThe guidance in NEI 99 01 is not intended to be applied to plants "as is"; however, developers should attempt tokeep their site specific schemes as close to the generic guidance as possible. The go.-al is to meet4 the intent of thgeerc nittiat-in-g Coniditions (IC-s) and Emer:.gency Actioni Levels (EA~s) withini the conlteEt of site speeificchara.cteA.....ri.stices locale, plant design. ope.ating terminology, etc. Meeting this goal will r.esult in ashorter and less cumffberso~me NRC review and approeval process, closer alignment with the schemes of othernuclear 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 beunambiguous and readily assessable because both serve specific purposes. As discu-ssed in Section 3, the generic9 .......R. 4;C ...... I _ ..e a..... oguidance-. inldssandd example EALs. it is the intent of thiis guidancee th~at both be included in site specificdocuments as eac. serves a specific purpose. The IC is the fundamental event or condition requiring adeclaration. The EAL(s) is the pre-determined threshold that defines when the IC is met. To this end, theSTPEGS ICs and EALs were developed with input from key stakeholders such as Operations, Training, HealthPhysics. and Engineering. STPEGS specific indications, parameters and values were consistent with licensingbasis documents, plant procedures, training, calculations. and drawings f some fetr of the p location ordesign 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 notapplicable to the plant, then that WC and/or- EAL need noet be inceluded in the sit spegeshemze. ini thsessdev~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 arepresented in Appendix A, Acronyms and Abbreviations. Those specific to STPEGS were included to beconsistent 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-specificdefinitions. These words and terms are identified by being set in all capital letters (i.e., ALL CAPS). Thedefinitions 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 dependingupon 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 witha 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-iedthrough for all of the associated 1G identifiers&.-R may be used in lieu of A-N4 may' be used 6in lieu of9For 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 presentationmethod (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 separateRecognition 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 EASateented. ThiS mateial is not required to be in the final energency classification scheme basis document..4.2 CRITICAL CHARACTERISTICSAs discussed abov~e, developers are encouraged to keep their site specific schemes as close to the geniericguidancee as possible. When crafting the scheme, developer: should satisPv themiselvesSTPE.GS ensured thatcertain critical characteristics haverbeen wereimet. These critical characteristics are listed below.The ICs, EALs, Operating Mode Applicability criteria, Notes and Basis information are consistent withindustry guidance; while the actual wording may be different from NEI 99-01 Revision 6, the classificationintent is maintained. With respect to Recognition Category F., a-the STPEGS scheme included a sei espeeillesc-hemel muIst inlu~kde some type3 of uLser-aid to facilitate timely and accurate classification of fission productbarrier losses and/or potential losses. The user-aid logic is-nmust-be consistent with the classification logicpresented in Section 9.*EAL statemnents use obljective criteria and observable values.* Cs.. EALs, Operating Mode Applicability and Note statements and formatting consider human factors andare 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 EALSlnStrum1entation referenced in EAL statements shudinelude that described in the. Pemergency plan section whichaddr-esses 10 CFR 50.47(b)(8) and (9) and/Or Chapter 7 of the FSAR. !nstrumentation utsed fer E;A~s need niet besafety 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 incorporatedinstrumentation 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 thedescribed event or condition.Scheme developers should ensure that specified values used as EAL setpoints are within the calibrated range ofthe referenced instrumentation, and consider any automatic instrumentation functions that may impact accurateEAL assessment. In addition., EAL setpoint values sheum4-do not use terms such as "off-scale low" or "off-scalehigh" since that type of reading may not be readily differentiated from an instrument failure. Findkhgs- andviolationis related to E ins. t rum....entationl may be locat-ad on the NR. website. If instrumentation failuresn. Fit i 0I .r A I c 00011, ot1A,it* t-n,, (1, ---oc --,1 tn, , aI~anirninco-~, f n n rimplementation may be used as described in plant procedures.;L- 4.4 PRESENTATION OF SCHEME INFOPAMATION TO USERSThe US Nuclear 'egulatoi' Comzm..ission. (NRC=) e..pects licensees to establish and maintain the capability toassess, eclassify and declare anl em .ergenc.y c.ndition prmptly within 15 minutes aer the availabilit ofinidicationis to plant operators that an emergeney, action level has been, or- may be, eEceeded. When wiin anemergency classification procere and creating r.elated user aids., the developer must determine the presentationmaethod(s) that best supports the enid users by, fac-ilitating accurfate and timely c mergency classification. To) thisend, developer-s shouild conisider the following, points.The first users ofan em.er.gency classification! afe the operators in the Contro.l Rom. During theae .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 tono~tice a change i pAn cn itins. They can assess the chianged conditions and, when w1am~anted, r-ecommenld adi:fferent e-m-l ier, en cassific-ation level to) the Teehnie-al Support Center (T-SC) and/or Em3ergency) OprtinFaeility (EOF). Emergency Directors in the T-SC- and/or EO.F will have moree appetlunity to focus on making anemergencey 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 licenfseadoperators are mROSt comfort1-14able. Developers V.ill neead to wok losely With repr-esentatives from the Operationsand Operationis Triainling Departmnents to develop readily usable and easily understood classificationi tools (e.g-., aproced~ire and relate uiser aids.). if necessary, an altern-atea mfethod for- prsetigemergency classific-ationschemfe information m~a" be developed for- use by Emiergency Directors anidor Offsite Response Organizationipe!-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 makea 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; ar-efer-ence to a Note is acceptable as long as the inform4:ationi is adequiately captur-ed on the wallboard and Pointedto by eacl applicable EAL,. Basis informa.io.nee not, be included an a A .allboad but it should be readilyavailable 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, startupand 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-eacceptable and include fieow charts, block diagramns, and celieklist type tables. Developer-s must ensurfe that thlesite specific method- ad-dre-sses all possible threshld cmbnaiosndclsic atio otcoeels Shown in theBAIR or PAIR EAL fission product barrier tables. The NRC staff considers the presentation m:ethod of theRecognition Category, F information to be an impoitanit user aid and may' request a change to a particutlarpro)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. TisLapproach would gr-eatly inresethe admfinlistrative cOntrols anid workload fer mnainitaining proeedures.On- theother hanld, performance ch-allenges mayf occur if recognitionl ofFmeeting -An IC- Orf EL4 i-S bazed SelelY, On tileffemom~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-ealerting the reoadder/uIster toe con;;s-Ul th~e site. emergency classific-ation procedur-e. Visual cues could be placed inemergency) operaltinig procedures, abnormlal operating procedures. alarm responise procedures, anid normaoperating procedures that apply to cold shutdown and r-efueling moades. As an exiample, a step, no~te- or cauMtioncould be placed at the beging of n CSlea abnorma o per-ating proeedurFe that rmnsthe rea;der that anemergency) classifiation assessmenit shouild be perbformed.4 6 -BASIS-DOCJLANTA basis docueknet is an integral part ofan emer-gency classification scemine. The material in this docuent~fesupports proper- emergency cl-assific-ationA deiio aking. by' prov'iding, iniforming backigro-und and developmentin4;fonnationf in a readily accessible formiat. it can be r-eferrFed to in training situations and wheni making ani actulalemergency classification, if necessary. The docue-ment is also; useful for establishing configuration managementcntrols for EP related equipmient and exiplaining.. aneergeny classification to off-site authorities. The contentoffthe 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 insection 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-edureEachi initiating Condition along with the associated EALs or- fission pr-oducet barrier thr-esholds, Operating ModeApplicability, Notes and Basis inforemation.o18 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 explainingor 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 rI-, 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. Ferexiample, a copy of the basis document could be maintained in the appropriate emergency r-esponse facilities.4.4 REFERENCES TO STPEGS AOPS AND EOPSAs reflected in the generic guidance, Some of the criteria/values used in several EALs and fission product barrierthresholds may-be were drawn from a p!a;44t'sSTPEGS AOPs and EOPs. This approach i-was intended tomaintain good alignment between operational diagnoses and emergency classification assessments. e-elepei-sshakild v'e'i4-4 th-at STPEGS verified the appropriate administrative controls are in place to ensure that asubsequent change to an AOP or EOP is screened to determine if an evaluation pursuant to 10 CFR 50.54(q) isrequired.s GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS-i5.1_GENERAL CONSIDERATIONSWhen making an emergency classification, the Emergency Director must consider all information having abearing 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 theRecognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholdsserve the same function as an EAL.NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare anemergency condition within 15 minutes after the availability of indications to plant operators that an emergencyaction level has been exceeded and to promptly declare the emergency condition as soon as possible followingidentification of the appropriate emergency classification level. The NRC staff has provided guidance onimplementing this requirement in NSIR/DPR-ISG-0 1, Interim Staff Guidance, Emergency Planning for NuclearPower Plants.All emergency classification assessments should be based upon *4a-aVALID indications, reports or conditions. Av#ald VALID indication, report, or condition, is one that has been verified through appropriate means such thatthere is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. Forexample, validation could be accomplished through an instrument channel check, response on related orredundant indicators, or direct observation by plant personnel. The validation of indications should be completedin 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 EmergencyDirector should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiologicalrelease is detected and the release start time is unknown, it should be assumed that the release duration specifiedin 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 notwarrant an emergency declaration provided that 1) the activity proceeds as planned and 2) the plant remainswithin 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 operatinglicense provided that the activity proceeds and concludes as expected. Events or conditions of this type may besubject 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 aspecific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak ratecalculation, etc.); the EAL and/or the associated basis discussion will identify the necessary analysis. In thesecases, the 15-minute declaration period starts with the availability of the analysis results that show the thresholdto be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees toestablish 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 maywarrant emergency classification, a provision for classification based on operator/management experience andjudgment is still necessary. The NEI 99 01 This scheme provides the Emergency Director with the ability toclassify events and conditions based upon judgment using EALs that are consistent with the EmergencyClassification Level (ECL) definitions (refer to Category H). The Emergency Director will need to determine ifthe effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. Asimilar provision is incorporated into the Fission Product Barrier Tables; judgment may be used to determine thestatus of a fission product barrier.5.2 CLASSIFICATION METHODOLOGYTo make an emergency classification, the user will compare an event or condition (i.e., the relevant plantindications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of anEAL(s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met orexceeded, 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 timeduration runs concurrently with the emergency classification process "clock." For a full discussion of this timingrequirement, refer to NSIR/DPR-ISG-01.5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONSWhen multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. Thehighest 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 atone 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 bedeclared.Related guidance concerning classification of rapidly escalating events or conditions is provided in RegulatoryIssue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During QuicklyChanging Events.5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATIONThe mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, isthe mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in amode change before the emergency is declared, the emergency classification level is still based on the mode thatexisted at the time that the event or condition was initiated (and not when it was declared). Once a different modeis reached, any new event or condition, not related to the original event or condition, requiring emergencyclassification should be evaluated against the ICs and EALs applicable to the operating mode at the time of thenew event or condition.For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the ColdShutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plantresponse. In particular, the fission product barrier EALs are applicable only to events that initiate in the HotShutdown mode or higher.5.5 CLASSIFICATION OF IMMINENT CONDITIONSAlthough EALs provide specific thresholds, the Emergency Director must remain alert to events or conditionsthat could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECLis IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergencyclassification should be made as if the EAL has been met. While applicable to all cmergency cla.sificatin4le4esEMERGENCY CLASSIFICATION LEVELS, this approach is particularly important at the higherem.ergency ela.sifieatien leve EMERGENCY CLASSIFICATION LEVELS since it provides additional timefor implementation of protective measures.21 IPaae 5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING ANDDOWNGRADINGAn 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, thenew 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 LongerExistsUnusual Event UNUSUAL EVENT Terminate the emergency in accordancewith plant procedures.A4e4ALERT Downgrade or terminate the emergency inaccordance with plant procedures.Site Area ...... egneySITE AREA Downgrade or terminate the emergency inEMERGENCY with no long-term plant accordance with plant procedures.damageSite Area AREA Terminate the emergency and enterEMERGENCY with long-term plant recovery in accordance with plantdamage procedures.Genera! E:mergencyGENERAL Terminate the emergency and enterEMERGENCY recovery in accordance with plantprocedures.As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS2007-02.2,65.7CLASSIFICATION OF SHORT-LIVED EVENTSAs discussed in Section 3.2, event-based ICs and EALs define a variety of specific occurrences that havepotential or actual safety significance. By their nature, some of these events may be short-lived and, thus., overbefore the emergency classification assessment can be completed. If an event occurs that meets or exceeds anEAL, 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 reactorfollowed by a successful manual scram/trip or an earthquake.5.8 CLASSIFICATION OF TRANSIENT CONDITIONSMany of the ICs and/or EALs contained in this document employ time-based criteria. These criteria will requirethat 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 anEAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance shouldbe 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 anexpected (normal) plant response, an emergency declaration is not warranted provided that associated systemsand 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 takesprompt manual action to address a condition, and the action is successful in correcting the condition prior to theemergency declaration, then the applicable EAL is not considered met and the associated emergency declarationis not required. For illustrative purposes, consider the following example.An ATWS occurs and the auxiliary feedwater system fails to automatically start. Steam generator levels rapidlydeerease lower and the plant enters an inadequate RCS heat removal condition (a potential loss of both the fuelclad and RCS barriers). If an operator manually starts the auxiliary feedwater system in accordance with an EOPstep and clears the inadequate RCS heat removal condition prior to an emergency declaration, then theclassification 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 obviatethe need to classify the event; emergency classification assessments must be deliberate and timely, with no unduedelays. The provision discussed above addresses only those rapidly evolving situations where an operator is ableto take a successful corrective action prior to the Emergency Director completing the review and steps necessaryto make the emergency declaration. This provision is included to ensure that any public protective actionsresulting from the emergency classification are truly warranted by the plant conditions.5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT ORCONDITIONIn some cases, an EAL may be met but the emergency classification was not made at the time of the event orcondition. This situation can occur when personnel discover that an event or condition existed which met anEAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. Thismay be due to the event or condition not being recognized at the time or an error that was made in the emergencyclassification process.In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 isapplicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 within onehour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State andlocal agencies in accordance with the agreed upon arrangements.5.10 RETRACTION OF AN EMERGENCY DECLARATIONGuidance 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/EALSTable AR-I: Recognition Category "AR" Initiating Condition MatrixUNUSUAL EVENTA-U4RUI Release ofgaseous or liquidradioactivity greaterthan 2 times the (siespecific effluent r-eleasecontrolling doeument)ODCM limits for 60minutes or longer.Op. Modes: AllA-IRU2UNPLANNED loss ofwater level aboveirradiated fuel.Op. Modes: AllALERTSITE AREAEMERGENCYGENERALEMERGENCYAA4RA1 Release ofgaseous or liquidradioactivity resulting inoffsite dose greater than10 mrenl TEDE or 50mrem thyfei4THYROIDCDE.Op. Modes: AllAA-RA2 Significantlowering of water levelabove, or damage to,irradiated fuel.Op. Alodes: AllA-A-3RA3 Radiationlevels that impedeaccess to equipmentnecessary for normalplant operations,cooldown or shutdown.Op. AM'odes: AllASIRSI Release ofgaseous radioactivityresulting ill offsite dosegreater than 100 mremTEDE or 500 mremthyro4iTHYROID CDE.Op. Modes: AllA2RS_2 Spent fuel poollevel at 40'-4" or lower(site speci: Ae.el 3dese~iptian4Op. M~odes: AllA-4RG_ Release ofgaseous radioactivityresulting in offsite dosegreater than 1,000 mremTEDE or 5,000 mremthyiredTHYROID CDE.Op. Modes: AllA-GRG.2 Spent fuelpool level cannot berestored to at least 40'-4'- (site specific Level 3des....p.iefi. for 60minutes or longer.Op. Modes: All241 P a e AUMRU1ECL: NetiieatiAPof Unusual Event UNUSUAL EVENTInitiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the (site speeifiecffluent release controlling document) ODCM limits for 60 minutes or longer.Operating Mode Applicability: A4IALLFefxF-Ape-Emergency Action Levels: (1 or 2 or 3)Notes:* The Emergency Director should declare the Unusual Event UNUSUAL EVENT promptly upon determiningthat 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 hasexceeded 60 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the releasepath, 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 specificefflueint release contro.lling, document) limits shown the values listed in Table RI column "UE" shew~for 60 minutes or longer:(site specific1I ....4monitor list and4 tl~rcs118dt Values ean-espn~ndan to 2 times the eentiRilln- lsmnlimits)Release PointUnit VentMain SteamLinesMonitorRT-801 OBRT-8046 thni8049-1-1.50 E+Table RI: Effluent MonitorsGE SAE ALERT08 ttCi/sec 1.50 E+07 utCi/sec 1.50 E+06 uCi/sec02 OCi/cni34.00 E+01 gCi/cm34.00_E+00 pCi/cm3UE1.40 E+05 tCi/sec5.00 E-02 LCi/C rn34.00 E+(2)Reading on .ANY effluent r-adiation monitor. g.eater than 2 times the alar setpoint established by a.....R on gaseous effluent radiationmonitor RT-8010B greater than 2 times the alarm setpoint established by a current radioactivitydischarge permit for 60 minutes or longer.(3) Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2times 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-levelradiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolledrelease). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those forwhich a radioactivity discharge permit is normally prepared.Nuclear pcwe" plants STPEGS incorporated design features intended to control the release of radioactiveeffluents to the environment. Further, there are administrative controls established to prevent unintentionalreleases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactiverelease 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 cannotbe readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant conditionand 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. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, thenthe 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 30minutes does not meet the EAL.EAL #1- This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous orliquid effluent pathways.EAL #2- This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2times the limit established by a radioactivity discharge permit. This EAL will typically be associated withplanned 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 orenvironmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into stormdrains, heat exchanger leakage in river water systems, etc.).Escalation of the emergency classification level would be via ICA-A4RA J.RUI: EAL-1 Selection BasisThe Unit Vent and Main Steam Line monitor readings were included in this EAL because they giveinstantaneous indications of a monitored gaseous release exceeding twice the ODCM limits. Normal gaseouseffluents are due to planned RCB purges and monitored by the Unit Vent. The Main Steam Line monitorreadings were included because they correspond to a concentration that Would result in a release rate of twice theODCM limits if there were a release via the Power Operated Relief Valves (PORVs) or Safety Relief Valves. Arelease from the PORVs or Safety Relief Valves is not a normal effluent pathway but meets the intent of theEAL.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 BasisFor EAL-2, there are two effluent radiation monitors, RT-8038 (liquid) and RT-8010B (gaseous). however onlyRT-8010B was included. The alarm setpoint for the gaseous effluent radiation monitor RT-8010B is set at theODCM limits. An indication of two times the alarm setpoint (two times the ODCM limit) would allow operatorstime to secure the release prior to meeting this EAL. The liquid effluent radiation monitor RT-8038 was notincluded in EAL-2 because the activity in liquid dischariges is normally the several orders of magnitude lowerthan the ODCM limits. In order to alert personnel to significant changes in the liquid effluent activity, the alarmsetpoint for RT-8038 is normally set several orders of magnitude below the ODCM limits. Setting the alarmsetpoint 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 EmergencyAction Levels2. Offsite Dose Calculation Manual (ODCM), Rev. 17. Part B3.0 to B4.93. 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 Man1ualThL11. ese Enocuments implement r-eguipuions reiaico to eT-luent controls, c.g.. Iu Q~ Fart -,w ang Lu v -lttPart 50, I). As appropriate, the RETS ..r COGrC- methodology should be .sed foF establishing themonitor 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 thatae no.t described in the RETS or OD.M..6. if.inluded, EAL values fo. these nit1orS Should be d.tes..inedusing the most applicable dosc,trelcasc iimis p-rs-nted in the RETS orF ODCM. it is rcoegnizced that a ealcuiatcdEAL value may be below wh.at the mo.nitor an Fad; in that case, the mniitor does not need to be included in thelist. Also, soen moiosmy notý be governed by Technicial Speciflcationis or other licenise related r-elatedrequiremenits; ther-efoare, it is impor'tant that the -a-s-soiated EAL and basis sete learly identify any limaitationsonH 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 timesrelease control limiit. The eontrollinig docuimenit typically describes methodologies fo9r deterinRing~l effluentradiationR monitor setpoints; these methodologies should be used to determine EAL .alues. in cases where amethodologyý is not adequately defined, develepe~s hould determ~ine valuies consistent with effluent conitrolregulations (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-mitto w.arn efa release that is not in compliance with the specified limits. indexing the value in this manner enstresconisistency) 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) therearc no au.tomati features that may render the monitor reading invalid (e.g., an auto purge feature triggered at apariciular indicationi level).it is r-ecognized that the condition descr-ibed by this IC- may result in a radiological effluent value beya'nd theoperating 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.nitorFeading minght be sei at 900% to 95,% of the highest accurate monitor i-adn. This prov~ision notwithstanding, ifthe 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 EALtlheshold.indications fromi a real time dose pricejetion system are not inlu~tded ini the genieric E.Ats. Many, liceense do. notha~ve this capability. For those that do., the capability may not be within the scope Ef Whipat TeehnicalSpecifications. 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 Hothave this capability. For- these that do, these monitor-s may not be controlled and maintained to the same le'vel asplant equipment, orf w~ithin the seope of the plant Technical Specifications. in addition, readings May beinfluenced by environm:eiital or- ethei factors. A licenfsee may r-equest to ineludc an EAL using a perimieternienitoring~ sytm prvlWill be CEnIISidered en a case by ease basis.ECL AssigajnmentAttributes: 3.l1.l1.B29 1 P a i4 e A2 RU2ECL: Netifieation of Uiusual Ev'ent UNUSUAL EVENTInitiating Condition: UNPLANNED loss of water level above irradiated fuel.Operating Mode Applicability: -I4ALLEkI 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 thefollowing:* Visual ObservationOR* 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 LevelHI/LO) is in alarmANDANDDin.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 onlyOR" RE-8099 (68' RCB) -Mode 5 or 6 onlyOR" 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 radiationlevels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in theability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety ofthe plant.A water level d-eerease-lowering will be primarily determined by indications from available levelinstrumentation. Other sources of level indications may include reports from plant personnel (e.g., from arefueling crew) or video camera observations ble). A significant drop in the water level may also causean 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 monitorreading may ineFeaserise due to planned evolutions such as lifting of the reactor vessel head or movement of afuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to anUNPLANNED loss of water level.A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance RecognitionCategory C during the Cold Shutdown and Refueling modes.Escalation of the emergency classifieation lec" EMERGENCY CLASSIFICATION LEVEL would be via ICAA-2RA2.RU2: EAL-1 Selection BasisHi/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 thenormal water level of 66'-6" (UFSAR 9.1.2.1), the "SFP WATER LEVEL HI/LO" lampbox 22M02 window F-5annunciator alarm is received in the Control Room (0POP09-AN-22M2. Annunciator Lampbox 22M02Response Instructions).Although the ICSA has a Hi/LO level sensor, there is not an annunciator in the Control Room similar to the onefor the Spent Fuel Pool. There is however, a "SFP TROUBLE" lampbox 22M02 window F-6 annunciator in thecontrol room. One of the inputs to this alarm is FC-LSHL-1420, the ICSA HI/LO level sensor. Since no fuel islocated 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 bioshieldwall close to the refueling cavity. Area radiation monitor RE-8090 is located in the Fuel Handling Building on68' 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 thetop of any array, the dose rate would be expected not to exceed 2.5 mR/hr, above background. This assumes 42hours 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 fromseveral hundred R/hr to over a thousand R/hr on and around the 68' elevation deck (table C-S NUREGCR/0649).311 Pa e
REFERENCES:
- 1. OPOP09-AN-22M2, Rev. 25, Annunciator Lampbox 22M02 Response Instructions F-5 and F-6Window (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 MonitorsDeveloper- Notes.:The "site specific lev~el indieatiens" are those inidications that may be used to moenitorf wAater level in the ya~elff1ions;; of the R-EFUELING PATHWAYA2. Specify the made applicability, of a pailiculaf indication if it is notavwailable in all moedes.The "site specific list of area radiation monitors"~ should contain toear-ea r-adiation monitor-s that would beexpected to have incr-eased readings following a decrease in water- lev~el in the site specific REFUELINGPcAsTHWJA'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/orstimplevels.Development Of the EALS shoulld- consid-er the -availability and limitations of mode dependent. Or otherc-ontrolled but temperaly,, r-adiation mionitors. Speeif.' the moede applicability of a pai~ieulaf monflitor if it is notavailable in all modes.iý-Al Ac-cza.111 A ,321Pai4e AAMRA1ECL: ,k!e4-ALERTInitiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10mrern TEDE oi 50 rnrem thyiad-THYROID CDE.Operating Mode Applicability: AIIALLxamfpe-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 applicabletime 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 hasexceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the releasepath, 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 classificationassessments 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 listedin Table RI column "ALERT" for 15 minutes or longer:(Site specific M...itor liSt and values)Table RI: Effluent MonitorsRelease Point Monitor GE SAE ALERT UEUnit 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/secMain Steam IT-8046 thru 4.00 E+02 uiCi/cn3 4.00 E+01 ýLCi/Crfl3 4.00 E+00 LICi/cm3 5.00 E-02 [1Ci/cmn3Lines 8049(2) Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mremthyfeid-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 dosesgreater than 10 mrem TEDE or 50 mrem thyfaid-THYROID CDE at or beyond (site specifi e4--sei'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 poinothe 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 onehour of inhalation.331 P age Basis:This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite dosesgreater than or equal to 1% of the EPA Proteetive Action GuidesPROTECTIVE ACTION GUIDES -(PAGs). Itincludes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potentialsubstantial degradation of the level of safety of the plant as indicated by a radiological release that significantlyexceeds regulatory limits (e.g., a significantuncontrolled release).Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannotbe readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant conditionand 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 wasestablished 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. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, thenthe effluent monitor reading is no longer a4,4diVALID for classification purposes.Escalation of the emergencyAS-I-RS 1.C Iacztiteation4eveI-EMERGENCY CLASSIFICATION LEVEL would be via ICRAI: EAL-1 Selection BasisThe Unit Vent and Main Steam Line monitor readings were included in this EAL because they giveinstantaneous indications of a monitored gaseous release meeting the EAL threshold values of 10 mnremTEDE or 50 mrem CDE THYROID at the SITE BOUNDARY. Gaseous releases from the plant aremonitored by the Unit Vent. The Main Steam Line monitor readings correspond to a concentration thatwould result in a release rate meeting the EAL threshold values if there were a release via the Power OperatedRelief 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 calculatedvalues to ensure they are readily assessable.RAI: EAL-2. 3. 4 Selection BasisN/AREFERENCES:1. Calculation No: STPNOCO13-CALC-002 Rev. 2.. Radiological Release Thresholds for EmergencyAction Levels2. 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 classificationdiversity aind maN, be bisee to elassify events tha weule not r-eaefn Vie saie +-l-, based on pliant statuls or tne341 P age fiSSionI product m-atrix, -Al]e For- many of the DR3A s anflalyz~ed in the UP dated Final Safety Analysis Report, thediSfrimfinlatOr VAIl not be0 thPe numfber- offission producet barriiers challenged, bult rather the amlOUnt Of rad ioactiyvityreeased to the environm~ent.The EPA PAGS arc expressed in tearms of the suim of the effective dose equivalent (EDE) and the comm~ittedeffective dose equivalent (CEDE), or as thie thyroid coEmmitted dose, equivalent (C-DE). For the pur-ose of theseIdEALs, 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 havedecided to base ptetie ations on child th)yroid GDE. Nuclear powver plant IGsiEAtLs need to be consistentwith the protec-ti've action mnethodologies employed by the States within their EPZs. The thyroid CDE dose utsedin the WC and EALs should be adjusted as necessary to aligni with State protectiv~e action decision makingThe "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 atthe "site specific dose r-eceptor- point" (conisistent with the calculation methodology employed) foar onie hourMoio 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 moitor4readings Bfr WsC and AG I. Acceptable sour.es of*this infrm.ation include, but are not limited to, theRETS/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 selectedmix should be the same as that employed to calculate Monitor readinigs fori W~s AS!I and AG 1. Acceptablesore f this; in;formation incjlude, but are-1 noAt limitead to, the RE2T-S/ODCN4 and vp-alues useLd in4 the. site'sLrem.r .e ...... A assessment methodology..Depending upon the methodology' used to calclate the EAL values, ther.e i..a:. be ov..erlap of some valuesbetween different Wds. Developer-s wvill nieed to address this over-lap by' adj usting these valuies in a m~annerthat 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 distiniguishbetv.'e~n on; site- and offsite doses. The selcte distance(s) aiid'or locations should r-eflect the content of theemferg~ency' plan, and the procedural methodology' used to deter-mine off-site doses and Proetective ActioniRecommendations. The variationl in selected dose recepto ponsmas there may be some differencees in thedistance 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-eare no automatic featurfes that may' render the m~onitori reading invalid (e.g., an auto purge feature triggerfed at -apariciular indicationi level).it is recogniz~ed that the condition described by this WC may resuilt inl a radiological effluent value beyond theoperating Or display' rang,=e of the installePd_ efflent moio.In those cases, EAL values should be determinedwith a margin sufficient to ensure that an accurfate monitor reading is available. For example, an EAL monitorr-eadinig might be set at 909% to 9-5%4 of the highest accurate monitorf re-ading. This provision notw~ithstanding. it35 1P a iz e theestmatd/clcuate moiteir reading is gr-eater than approximately 1109% of the highest aecur'ate iiion itorreading, then developers may sli dse not to include the monitor as an indieation mid identity an alterate EALdifeshekk.~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 nothave this capability. For those that do, the eapability may not be within the scope of the plant TFechnicr-alSpecifications. A licensee may r~equest to include an EAL using real time dose proejecto ytmrslsapproval 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 nothave this capability. F-Or those that do, thiese mnonitorFs may not be controlled and- maint~ained to the same level asplant equipment, or withini the scope o~f the plant Technicial Specifiations. In additioni, r-eadinigs miay-4beinfluenced 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 AssignglmentAt:ri4hit;"A I 1 I -, --36 1 P ag e AA2RA2ECL: Alei4-ALERTInitiating Condition: Significant lowering of water level above, or damage to, irradiated fuel.Operating Mode Applicability: A-gALLExample 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 thefollowing FHB radiation monitors readings:* FHB Exhaust, RT-8035 or RT-8036 greater than 1.00 E- I LtCi/crn3OR* ARM (68' FHB), RE-8090 greater than 1.500 mR/hrORb. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of thefollowing RCB radiation monitor readings (Mode 5 or 6 only).0 ARMs (68' RCB), RE-8055 or RE-8099 re'eater than 850 mR/hr.NOTEEAL-3 is 17ot applicable u1ntil the enhanced SFPlevel instrumentation is available for use.(3) Lowering of spent fuel pool level to (sitelower.po,,ifie 2 value). [See Dev'lopior 49'- 10" orBasis:-This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or asignificant lowering of water level within the spent fuel pool (icc Develpeir Notes) or Inside ContainmentStorage Area (ICSA).-These events present radiological safety challenges to plant personnel and are precursors toa release of radioactivity to the environment. As such, they represent an actual or potential substantialdegradation 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 issealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified inaccordance 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 REFUELINGPATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiatedfuel 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 mayalso be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality ofavailable indications, reports and observations. While an area radiation monitor could detect an in adose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may notbe a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings shouldbe 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 RecognitionCategory 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 ontoan assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reportsor 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 preventsignificant dose consequences from direct gamma radiation to personnel performing operations in the vicinity ofthe spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is alsoa 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 ICsAS-1-RS I or A8-RS2 (sc 1S2 A T , -:;'sRA2: EAL-2 Selection Basis:The calculated airborne source term and radiation monitor responses for a fuel handling accident in the FHB isbased on Calculation STPNOC013-CALC-005 Rev. 1. The threshold value of 1500 inkhr for area radiationmonitor RE-8090 was truncated less than 4% from the calculated value to ensure the threshold was readilyassessable. Threshold values for FHB Exhaust Monitors RT-8035 and RT-8036 were also included because theyare accident monitors that are sensitive to noble Pases which are expected to be present if irradiated fuel isdamaged. The calculated monitor reading for RT-8035 and RT-8036 is 3.8 uiCi/cm' and the high range of themonitors is 0.3 utCi/cm-. The threshold value of 0.1 ,.Ci/cmn is approximately 6 orders of macnitude abovebackground and indicative of damaged irradiated fuel. It was selected because it is readily assessable and withinthe calibrated range of the monitors.The calculated airborne source term and radiation monitor response for a fuel handling accident in the RCB isbased 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 Regardto 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 substantialradiation 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 12-02indicates 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 2criteria as 49'- 10".REFERENCES:1. Calculation No.: STPNOCO I 3-CALC-005 Rev. 1, Fuel Handling Accident Monitor Response forEmergency 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 FuelPool Level Instrumentation to Meet NRC Order EA-12-051, Rev. 0, NOC -AE-13002959Developer- Notes:FerE~AL #1Depending upon the availability and range of inistrumientation. this EAL may include specific readings indicativeof4f@e uo eeR;cnSider wAtelad- r-adiatiOnlevl tOE4readings. Speeify, the made applicability 4fa particularindircation; if it ik not available in All m;AdeFor EAL #2The "4site specific listing of r-adiation monitor-s, and the assocei-ated- reaadings. setpoints and/lor a~alarms" shouildcontain 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-) thereae no automatic featur-es that mfa:, render the monitor r-eading invalid (e.g., an auto purg-e featur-e tr-iggered ataparticular 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," OFdisplay ranige of the inistalled radiation monitor-. hin those eases, EAL vausshould be deemndwithamrisufflieent to ensurc that an aeeurate mionitor, reading is ava~iable. For examnple, an EAL mnonitor r-eading might beset at 90% to 95-% of the highest accurate moit)Ror readinkg. This provision notwithstanding, if theestimated/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 EALthireshe14.To further promote aeecurate classification, developer-s sheuld consider if same combination of monitors could beclassification 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 otherconHtrolled but temipEoraf~'. Fadiation moneiitor-s. Specify the mode applicability of a paricuelar- monitorF if it is notavaPi]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 beimplemented whben the enhanced spent fbel pool level instrumwentation is available for- use. The "site specificLevel 2 value" is uisually, the spent fuiel peal level tha! is adequate to provide substantial radiation shielding for aperson standing on the spent fuiel pool operating deck. This site specific level is determined in aecofdanc-e withNRC 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:itationisassciaed 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.G40 1 P a R e AA3RA3ECL: Alei4 ALERTInitiating 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 eventoccurred, 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 thefollowing plant racms. or areas listed in Table H3/R2:TABLE H3/R2: Plant Areas Requiring AccessRCB RHR Heat Exchanger Roomso *1 MAB 51 ft Room 335EAR Roof, MCC 1G8, 4.16KV Switchgear Roomso "nl EAB 4.16KV Switchgear Rooms41 1Pa e (site specifi list of planit rEms or areas with entry related mode applieability identified)Turbine Genierator BEuildin (TGB)lselation Valve-Cbee(VFuel aninBilig(Fl-B)Reacter Containment Building (CBBasis:This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impedepersonnel from performing actions necessary to maintain normal plant operation, or to perform a normal plantcooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level ofsafety of the plant. The Emergency Director should consider the cause of the radiation levelsand 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. Theemergency classification is not contingent upon whether entry is actually necessary at the time of the iner-easedhigher radiation levels. Access should be considered as impeded if extraordinary measures are necessary tofacilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use ofnon-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., entryis not required during the operating mode in effect at the time of the elevated radiation levels). Forexample, the plant is in Mode I when the radiation i-nefeaserise occurs, and the procedures used fornormal 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 compensatorymeasures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter orresin 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 actuallyprevent or impede a required action.Escalation of the emergency classification level EMERGENCY CLASSIFICATION LEVEL would be viaRecognition 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 forexpected occupancy times. The rooms listed in EAL-1 require continuous occupancy to maintain normal plantoneration. 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, orshutdown. Assuming all plant equipment is operating as designed. Normal operations and safe shutdownequipment operation is capable from the Main Control Room (MCR). Tile plant is able to transition into a hotshutdown firom the MCR, therefore H3/R2 is a list of plant rooms or areas with entry-related mode applicabilitythat 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, andlimit dilution pathways) and subsequent entry into cold shutdown (disable operation of ECCS equipment). Afterachieving cold shutdown it is assumed that the plant will be maintained in a cold shutdown condition.REFERENCES:1. General Design Criteria 192. OPOP03-ZG-0008, Rev. 56. Power Operations3. OPOPO3-ZG-0006. Rev. 54. Plant Shutdown from 100% to Hot Standby4. OPOP03-ZG-0007. Rev. 71, Plant CooldownDeveepF NetesiEAL 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 eXpectedeceupan y tmsThe ....ther Site speeific arieas.rooms" should in..lude any areas or rooms requiring contin-ous occupancy tomIaint-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 specifythose rooms or- areas; that contain equipment whichl require a manual/loc-al action as spccified in operatingprocedures used for- normal planit oper'ationi, eooklown and shutdown. Do,niot include roams Or areas ill whic-hactions of a contingent Or emerg-ency natur-e wouild be perfor-med. (e.g., an action to address an cff normal oremergency, eonditiosi suchi as emergency repairs, correctiv~e measurfes Or emer-gency operations). in addition, thelist 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 anadministrative 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 allowedby 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.-AaECL Assignment Attributes: 3.1.2.C NEI 99 01 (.Revision 6) Novemiber 2012 1143 P a R e ASI-RS1ECL: Site Area Emergency SITESITE AREA EMERGENCYInitiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 rnrem TEDE or500 mremthyroid THYROID CDE.Operating Mode Applicability: A4UALLExam pie Emergency Action Levels: (I or 2 or 3)Notes:-The Emergency Director should declare the Site Area EmergencySITE AREA EMERGENCY promptlyupon 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 hasexceeded 15 minutes.e If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the releasepath, 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 classificationassessments 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 TableRI column "SAE" shown for 15 minutes or longer:(site speeifie monitor- list and thrfeshold valuesoTable RI: Effluent MonitorsRelease Point Monitor GE SAE ALERT UEUnit 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/secMain Steam RT-8046 tirt 4.00 E+02 itCi/cm3 4.00 E+01 yCi/cm3 4.00 E+00 ktCi/cm3 5.00 E-02 [tCi/cmn3Lines 8049__.___(2) Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mremthyfeidTHYROID 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 onehour 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 orequal to 10% of the EPA ProtecetiveAetion Guides PROTECTIVE ACTION GUIDES (PAGs). It includes bothmonitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systemsneeded for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannotbe readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant conditionand 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 wasestablished 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. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, thenthe effluent monitor reading is no longer valid -VALIDVALID for classification purposes.Escalation of the1.emerzenc;v celassiatiton level EMERGENCY CLASSIFICATION LEVEL would be via ICRSI: EAL-1 Selection Basis:The Unit Vent and Main Steam Line monitor readings were included in this EAL because they giveinstantaneous indications of a monitored gaseous release meeting the EAL threshold values of 100 mremTEDE or 500 mrem CDE THYROID at the SITE BOUNDARY. Gaseous releases from the plant aremonitored by the Unit Vent. The Main Steam Line monitor readings correspond to a concentration thatwould result in a release rate meeting the EAL threshold values if there were a release via the Power OperatedRelief Valves (PORVs) or Safety Relief Valves.The Unit Vent and Main Steam Line release values are based on Calculation No. STPNOCO13-CALC-002Rev.2. The adiusted values used in this EAL were conservatively truncated by less than 1% of the calculatedvalues to ensure they are readily assessable.RSI: EAL-2. EAL-3 Selection Basis:N/AREFERENCES:1. Calculation No: STPNOC013-CALC-002 Rev.2, Radiological Release Thresholds for EmergencvAction Levels2. 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 elassirieatiindiversity 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, the451 P a g e discime-oin;ator; will not bpe the. numbier of fission producet barr-iers challeanged, but r-ather- thea Amount of radioactivityFrele,-kaed- to- the evrnetThe EPA PAGs are expressed in terms of the sumi of the effeetive dose equivalent (EDE) and the committedeffctie dseequivalenit (CEDE), or as the thyroid committed dose equivalent (C-DE). For the purpose of thesIC/EAkks. the dose quantity total effective dseuvAln (T-EDE), as defined in 10 CPR § 20, is usead in lieu ofAe.ii f EDE andCE .The EPA PAG guidance proevides for the uise of adult thyr-oid dose conversion factors; however, some states hav.edcddto b-ase proatective ac-tions oni child thyr'oid C-DE. NJucleaF power plant !C~s/Elg~s need to) be conlsistentwith the proetective action methodologies em ple.,ed by the States within their- EPZs. The thyrFoid C-DE dose usedin the IC and EAL~s should be adjusted as necessar-y to align with State protective action decision maki-ngThe "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~idCDE; 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" atmohericdispersion fac-tors; the data or factor-s selected forf use should be the sam~e as those employed to calcuilatethe monitor- readinggs for !Gs AA 1 and AG 1. Akcceptable sou-ces of+this information include, but are notlimited 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 selectedmix should be the same as that employed to calculate mon1f.itor. r..adi..gs fr- WTs AA 1 and AGA.A 1sources of this hIIn matLI I bLit are not limited E a, the RETS/ODCM a'nd values usedin thp. 4pt'A; sourpres'; d.ow;.p .s ,. msthdlov.-1Depending upon the jnethodology, used to calculate the EAL values, there may be overlap of some valuiesbetween differ-ent W~s. Developers will need to addr-ess this overlap by adjusting these values in a manner- thatensur~es a logical escalation in the ECL.The "site speehifi e---- receRptor point" is the distance(s) andor locations used by the lieenseeto distinguisThbetween on site an;d- offsite doses. The selected distancee(s) an~bJor locationwSs shoulmId- reflect the conitenit of theemergenoy, plan, anld the proce~dural methodology used to determinife offSitea doescu And Proatective ActoRecommendations. The variation in selected dose recepto ponsmas there may, be som~e differences in thedist-ance froam the release point to the c Alculated dose point fr-omf site tosieDevelopers 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) thereare no auitomatic featur-es that mnay r-ender- the moenitor- readinig invalid (e.g., an auto pafrge feature trigger-ed at apar-ticularindcto level).it is recgnze tht h cnition descPribed by, this WC may, r-esult in a radiologial eff~luent v'llalu beyond theoperating or display range o~f the installed eft44fent moitor. fInthse c-ases. EAL values shouild be determ~inedwith 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,",, ithe estimated/calclated monitor reading is greater than apprOXimatel' 11090; of the highest accurate monfitor46 1 P a ge redig then devlapr ma hos o t nluete .oio as an. indicatiion mnd identify, an alter-nate EALthreshold~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 donot. ave hIs -capability. F-Or those that do, the eapabilit,' may, not be within the scope of the plant TechnicalSpecifications. 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~dications from a per-imetcrmntrigsse are fiOtincluded in the generic EALs. Many lieen:seesdE- Retha'.'eimay HOt fie CONIFOlled and maintailied te she s-ame level asplanlt eqUipmient, orf within the scope of the plant Technical Specifications. in additiotn, readings miay beinfluenced 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 .C471 P a e AS9RS2ECL: Site A:"ea Emergency SITE AREA EMERGENCYInitiating Condition: Spent fuel pool level at (site .peeifie Level 3 de.c.-iptin) lower.Operating Mode Applicability: AI4ALLE-taniple Emergency Action Levels:NOTEEAL-1 is not applicable until the enhanced SFPlevel 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 toIMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of thepublic 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 SITESITEAREA EMERGENCY IC was met; however, it is included to provide classification diversity.Escalation of the emergency elassifieatien leve- EMERGENCY CLASSIFICATION LEVEL would be via ICAG-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 Regardto 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 actionsto 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 Level3 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 ModifyLicenses with Regard to Reliable Spent Fuel Pool Instrumentation", August 20124. South Texas Proiect (STP) Overall Integrated Plan for Implementation of Unit I & Unit 2 Spent FuelPool Level Instrumentation to Meet NRC Order EA-12-051, Rev. 0, NOC -AE-13002959Developer Notes:hi aeear-dance with the discussion ini Section 1.1. NRC Order- EA 12 05 1, it is r-ecommended that this WC andEAL be implemented when the- enh11anceRd- Spent fuel pool level-0 ins.,-t.rumen;4.--;;ta;-tion-4 is lavailaible. for. use. The '"sitespecific Leel 33 valu'il" is us.ll:,' that spent fu.el pool level where fuel r ,emains ..vered and aetions to implementmake up water addition should no longer be deferred. This site specific level is determiined in accordance withNRC 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) refletassciated with the design or operation of instrumentation used toaRsite speeicife on4str-air#tS or limitationskiP+pR!!fl:lp- +QP f. , -".,Pi .VAI-P-rCT A ch: -I x,. I 21491 P a ge AG4RG1ECL: General EmergeneyGENERAL EMERGENCYInitiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDEor 5,000 mrent-hyre44 THYROID CDE.Operating Mode Applicability: AIIALLlAi-Fmple-Emergency Action Levels: (I or 2 or 3)Notes:* The Emergency Director should declare the General Emergeency GENERAL EMERGENCY promptly upondetermining 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 hasexceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the releasepath, 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 classificationassessments 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 TableRI column "GE" reading .h.on, .for 15 minutes or longer:(sthe speeifie moenitor lit and thresholdv;alues)Table RI: Effluent MonitorsRelease Point Monitor GE SAE ALERT UIEUnit 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/secMain Steam RT-8046 thru 4.00 E+02 LOi/cmn3 4.00 E+01 tCi/cm3 4.00 E+00 ICi/crn3 5.00 E-02 itCi/crn3Lines 8049(2) Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000mrem 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 onehour 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 orequal to the EPA Pretective Action Guides PROTECTIVE ACTION GUIDES (PAGs). It includes bothmonitored and un-monitored releases. Releases of this magnitude will require implementation of protectiveactions for the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannotbe readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant conditionand 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 wasestablished 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. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, thenthe 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 giveinstantaneous indications of a monitored gaseous release meeting the EAL threshold values of 1000 mrem TEDEor 5000 mrnrm CDE THYROID at the SITE BOUNDARY. Gaseous releases from the plant are monitored by theUnit Vent. The Main Steam Line monitor readings correspond to a concentration that would result in a releaserate meeting the EAL threshold values if the release was via the Power Operated Relief Valves (PORVs) orSafety Relief Valves.The Unit Vent and Main Steam Line release values are based on Calculation No. STPNOCOI3-CALC-002Rev.2. The adjusted values used in this EAL were conservatively truncated by less than 1% of the calculatedvalues to ensure they are readily assessable.RGI: EAL-2, EAL-3 Selection Basis:N/AREFERENCES:I. Calculation No: STPNOCO 1 3-CALC-002 Rev.2, Radiological Release Thresholds for EmergencyAction 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 AG2RG2ECL: General Emergeci'yGENERAL EMERGENCYInitiating 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: AIIALLExample Emergency Action Levels:Note: The Emergency Director should declare the General En'eigeneyGENERAL EMERGENCY promptlyupon determining that 60 minutes has been exceeded, or will likely be exceeded.NOTEEAL-1 is not applicable until the enhanced SFPlevel 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 minutesor longer.Basis:This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to aprolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to theenvironment.It is recognized that this IC would likely not be met until well after another General Emergnc:'y GENERALEMERGENCY 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 withRegard 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 actionsto 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 3criteria 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 ModifyLicenses with Regard to Reliable Spent Fuel Pool Instrumentation", August 20124. South Texas Project (STP) Overall Integrated Plan for Implementation of Unit I & Unit 2 Spent FuelPool Level Instrumentation to Meet NRC Order EA-12-051, Rev. 0, NOC.-AE-13002959Dev.eloperiNotes.in aeccrdance with the discussion in SeItion 1.4, NRC Order EA 12 05 1, it is recommended that this Ic andEAL be implemented when the enhanced spent fuel pool level instrumentation is available for uise. The "sitespec-ific Level 3 value" is usually that spent fuiel peel level wher-e fuel r-emainis covered -and atnsto implemenitmfake Lip water addition should no longer- be deferred. This site specific level is deter~mfined in accor-dance withO.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 limiitationsassociated with the design or oper-ation of instrumentation used to determnine the Level 3 value.ECL AssignmentAttrhi~~~1 1 4lfG53 1Page 7 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONICS/EALSTable C-i: Recognition Category "C" Initiating Condition MatrixUNUSUAL EVENTCUI UNPLANNEDloss of (Feaet-v~esse1,'RCS [PUW] i-eR4 [B;MR) r inventoryfor 15 minutes orlonger.Op. Modes: 5.6C-6/4Shitkldwii, 5n6ALERTSITE AREAEMERGENCYGENERALEMERGENCYCAI Loss of --eaete+vessel RCS [4WR-] er-inventory.Op. Modes: 5.6 Ce4/dCS1 Loss of (+feaet-vessel RCS -rJ% te.RPV -.B4R4.) inventoryaffecting core decayheat removal capability.Op. Modes: 5_6C-ldShutdoqwi, Refuoelhng 5,6CG1 Loss of (eaetefvessel RCS [Pff;R] or RPV[-WR]) inventoryaffecting fuel clad integritywith containmentchallenged.Op. Modes: 5,6C-mShtidow;'n, Refiuolig 5,6CU2 Loss of aI4ALLbut one AC powersource to emergencybuses for 15 minutes orlonger.Op. Modes: .56,6CdGShfikdown, Refuoling"-5-DefueledCU3 UNPLANNEDineefeaserise in RCStemperature.Op. Modes: 5.6C-e-dShutdown, Rei4ding 5,ý6CU4 Loss of Vital DCpower for 15 minutes orlonger.Op. Modes: 5.6C-oekS/htdo'wn, Ro.,,ing 5,6CA2 Loss of a#ALLoffsite and aI4ALLonsite AC power toemergency buses for 15minutes or longer.Op. Modes: 5,6 C-64dShuitdown, R.f. ,;ing-5,, DefuteledCA3 Inability tomaintain the plant incold shutdown.Op. M'odes: 5.6C-44RDf,.,lIg 5ý6CU5 Loss of allALLonsite or offsitecommunicationscapabilities.Op. Modes: 5,6. G-445--DefueledCA6 Hazardous eventaffecting a SAFETYSYSTEM needed forthe current operating541 P a e mode.Op. Modes. 5_6 C-44Shz~t~deivi, Refiieing 5,655 1 P a g e culMLA/ oti:,icaiI'cn Cf Unusual Event UNUSUAL EVENTInitiating Condition: UNPLANNED loss of(reaeter "esse!/RCS 'DPWR] , r RPV [BWR)] inventory for 15minutes or longer.Operating Mode Applicability: Cold Shutdown, Refucling 5,6Example Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the Unusual EventUNUSUAL EVENT promptly upondetennining 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 keeethan a r.equired lower lim.it below the Reactor Vessel Flange procedurally required limit for 15 minutesor longer.(2) a. Reaeter 'el/CS [PWR] or RP rWkJ.]) level cannot be monitored.ANDb. UNPLANNED inefeaserise in (SITE SPECIFIC SUMP AND/OR TANK) ANYANY of thefollowing 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 SumpBasis:This IC addresses the inability to restore and maintain water level to a required minimum level (or the lowerlimit of a level band), or a loss of the ability to monitor (reaeter -essel/RCS [PJJ'R] or RPV [8WIR]) levelconcurrent with indications of coolant leakage. Either of these conditions is considered to be a potentialdegradation of the level of safety of the plant.Refueling evolutions that deereaselower RCS water inventory are carefully planned and controlled. AnUNPLANNED event that results in water level decreasing below a procedurally required limit warrants thedeclaration of an Unusual Event UNUSUAL EVENT due to the reduced water inventory that is available to keepthe core covered.56 1P a g e EAL #1- recognizes that the minimumn required ([reaet. ,i vessW rC " o["! @!- RPV [BURj) level can changeseveral times during the course of a refueling outage as different plant configurations and system lineups areimplemented. This EAL is met if the mininmum1 level, specified for the current plant conditions, cannot bemaintained for 15 minutes or longer. The minimum level is typieally typieally specified in the applicable STPoperating 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 theexpected 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) levelhave been lost. In this condition, operators may determine that an inventory loss is occurring by observingchanges in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potentialsources of water flow to ensure they are indicative of leakage from the (reaetor vessel/RCS rP-WR] OF -fPVContinued loss of RCS inventory may result in escalation to the Ale'4-ALERT emergenicy classification le-elEMERGENCY 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 flangeduring refueling outages Ptper OPOP03-ZG-0007, Plant Cooldown.. RCS level below the ;essel flangeis to be minimized. RCS level may be dr-opnjglowered below the vessel flange for specific purposes *-rdeicontrolled evolutions (e.g.. head removal, mid-loop operations) as described in OPOPO3-ZG-0009, Mid-LoopOperation. The 15 minute time frame allows for prompt operator actions to restore RCS level in the event of anUNPLANNED 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 secondcondition 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. ExcessiveRCS Leakage. Since other system leaks could rise levels in various tanks and sumnpl)S. the list has been limited tothe tanks and sumps that would have the highest probability of indicating RCS leakage inside the ReactorContainment 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 Leakage2. OPOP03-ZG-0007, Rev. 71, Plant Cooldown3. OPOP03-ZG-0009, Rev. 59. Mid-Loop OperationDevlper- Notes!EAL -1 It is recognized that the minim.um allowable r1eactor..-.esse/R.S/.RPV level may have many valuesover. the course of a refueling outage. Developers should solicit input f..o. licensed perators concerning theoptimum wor-ding for this EAL statement. hi particular-, determine if the geei oding is adequate to ensur~e57 1 P a g e acuawie and timely classificatiten. E if speoif; setpoints. .an hbe. inc.uded without m.aking. the EAL stateme1nunw.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. eincluded, these should be dr-awn fromff applic-able operating, proedwre" et-othei- eentfolhing dacumensEAL 42.b Enter any "site specific sump and/or tank" levels that c:uld be expected to inrease ifthere werealoss of invento:ry (i.e., the lost inventory w"oulde R t L. Fthle listed Sump a!. otank).A "I .1 1 A1-44 ai~ssieimrnent Altrietites: 1.1..581Page CU2ECL: Nctificatien of Unusual Event UNUSUAL EVENTInitiating 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 upondetennining 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 isreduced to a single power source for 15 minutes or longer.ANDb. ANY additional single power source failure will result in loss of OIALL AC power to SAFETYSYSTEMS.Basis:This IC describes a significant degradation of offsite and onsite AC power sources such that any additional singlefailure would result in a loss of allAC power to SAFETY SYSTEMS. In this condition, the sole AC powersource 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 ALERTbecause of the incFteasedadditional time available to restore another power source to service. Additional time isavailable due to the reduced core decay heat load, and the lower temperatures and pressures in various plantsystems. Thus, when in these modes, this condition is considered to be a potential degradation of the level ofsafety of the plant.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power toan 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 onsitediesel generator).e A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with asingle 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 busesbeing 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 inaccordance 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 anyadditional single failure would results in a loss of all AC power. This condition is an UNUSUAL EVENTduring modes 5. 6 and Defueled because of the additional time available to restore power due to the reduced coredecay heat load, and the lower temperatures and pressures in various plant systems. In modes 1-4, this conditionis 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 Bus2. OPOP04-AE-0004, Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus3. OPSP03-EA-0002, Rev. 32, ESF Power Availability4. Drawing OOOOOEOAAAA. Rev. 24, Single Line Diagram, Main One Line Diagram, Unit No. I & 260 1 P age EDeN'e!E)e EAL-NeSiFor 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 sizedtoA fiee-d I A.C emer-gency bus), the EAL and Basis section must speei4ý that both generatorS fcr that source ar~eeeiang7The "site specific emergency5 buses" are the buises fed by off-site or- emergcney ~AG power sourcees that supplypowAer to the electric-al idistribuition system that powers SAFETY SYSTEMS. Theare, is typically 1 emnergency busper: train of SAFETY SYSTEMS.Developers should modify, the buileted examples provided in the beasis section. above, as nede t -fleet theirsite specific pln des Igns and capabilities.0The 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'atepowAer 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 etthis source is recognized in AOPs and EOPS.ý or beyond design basis accident response aguidelines (e.g-., FLEXStippei4 guidelinies). Suceh power: sources should generally m~eet the "Alterniate acG sourcFe" definition provided in10 C FR 50.2.~At multi unit stations, the EALs may, cr-edit opnstr measures that ar-e procedur-alized an;;d- can beimplemented within 15 minut-- es.Cosider capabilitikees suhas power source croess ties, "swing" genierators, othepower sources described in -abnonm.al or emergency operating procedures, etc. Plants that hav~e a proceduralizedcapability to supply off-site AC pov.'er to an aff-ected unit via a cross tie to) a comI~panionuit acei thispower 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.A61 1Page CU3ECL: Notification of Unusual Event UNUSUAL EVENTInitiating Condition: UNPLANNED inefeaserise in RCS temperature.Operating Mode Applicability: Cold Shutdown, Refueling 5. 6Example Emergency Action Levels: (I or 2)Note: The Emergency Director should declare the Unusual Event UNUSUAL EVENT promptly upondetermining that 15 minutes has been exceeded, or will likely be exceeded.(1) UNPLANNED i-neFeaserise in RCS temperature to greater than (site specific Technical Specificationcod szhutdown temperatur. e limit) 200 'F (Tavg).(2) Loss of ALL RCS temperature and (Feaete: HvesselRCS [.PR] or RPM [BWR]) level indication for 15minutes or longer.Basis:This IC addresses an UNPLANNED in+e-easerise in RCS temperature above the Technical Specification coldshutdown temperature limit, or the inability to deterniine RCS temperature and level, represents a potentialdegradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is notestablished during this event, the Emergency Director should also refer to IC CA3.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limitwhen 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 thatwhich can currently be removed, such that reactor coolant temperature cannot be maintained below the coldshutdown temperature limit specified in Technical Specifications. During this condition, there is no immediatethreat 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 coolanttemperature depending on the time after shutdown.EAL #2- reflects a condition where there has been a significant loss of instrumentation capability necessary tomonitor RCS conditions and operators would be unable to monitor key parameters necessary to assure coredecay heat removal. During this condition, there is no immediate threat of fuel damage because the core decayheat 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 toAlei4-ALERT would be via IC CAl based on an inventory loss or IC CA3 based on exceeding plantconfiguration-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 theinability to control RCS temperature. Mode 4 (Hot Shutdown) would be entered when Tavg exceeds 200 OF(Reference 1).CU3: EAL-2 Selection Basis:N/AREFERENCES: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 CU4ECL: Netimfcatien of Unusual Event UNUSUAL EVENTInitiating Condition: Loss of Vital DC power for 15 minutes or longer.Operating Mode Applicability: Cold Shutdown. Refieling 5. 6Exa*imple.Emergency Action Levels:Note: The Emergency Director should declare the Unustal E-vent UNUSUAL EVENT promptly upondetenrmining 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 for15 minutes or longer.Basis:This IC addresses a loss of Vital DC power which compromises the ability to monitor and control operableSAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decayheat load has been significantly reduced, and coolant system temperatures and pressures are lower; theseconditions inereaseextend the time available to restore a vital DC bus to service. Thus, this condition isconsidered 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, oroperable, 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 DCpower affecting Train B would require the declaration of an Unusual Event UNUSUAL EVENT. A loss of VitalDC 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 CLASSIFICATIONLEVEL 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 tobe 105.5 volts. At 105.5 volts or less, OPOP05-EO-ECOO, Loss of All AC Power, directs the operators to openthe battery output breakers.REFERENCES:1. Calculation I 3-DJ-006, Rev. 0. 125 VDC Battery Four Hour Coping Analysis2. OPOP05-EO-ECOO, Rev. 23. Loss of All AC Power64 1 P ag e Developer- Notcs:The "site spe;ifie bas oEtlytage value" should be based on the ..inimum b. svoltage neessary f.r adequtateOperation of SAFETY SYSTEM equipmfent. This voltage v~aluie s houLld incOr-porate a margin of at least 1 5minutes of operationi befor~e the oniset of iinabilit' to opefate these leads. This voltage is usually ncar hminimuim 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 eellvoltagge is approximately 1.75 Volts per cell. For a 58 string battery set, the minimum voltage is approxEimately1.8-1 Volts per cell!.ECL Assignment Attributes: 3.1..1At65 1 P a g e CU5ECL: Notification of Unusual Event UNUSUAL EVENTInitiating Condition: Loss ofa44ALL onsite or offsite communications capabilities.Operating Mode Applicability: C-ld Refueliiný-5, DefileledFxaniple 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 TableC4.(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-3ONSITE ORO NRCPlant PA system XPlant Radios XPlant telephone system X X XSatellite phones X XDirect line from Control Rooms to Bay City X XMicrowave Lines to Houston X XSecurity radio to Matagorda County XDedicated Ring-down lines XENS line XBasis:This IC addresses a significant loss of on-site or offsite communications capabilities. While not a directchallenge 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 multipleradio 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 emergencydeclaration. The OROs referred to here are (se De;'lepe: Notes) Matagorda County Sheriff's Office, and TexasDepartment of Public Safety Disaster District in Pierce.EAL #3-addresses a total loss of the communications methods used to notify the NRC of an emergencydeclaration.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 ofinformation. Links were obtained from procedures OPGP05-ZV-001 1, Emergency Communications.REFERENCES:1. OPGP05-ZV-00 11, Rev. 8, Emergency CommunicationsDeve'loper- Notes:1F=A -.41 The "site speeific list of com,,iatio methods" should include all ;ommunications meth.As usedfor rutine planlt commiunications (e.g., mercial or site telephones, page pa.. y syste..:s, radios. etc.). Thislisting should icueinsctalled plant equipment and components, and not itmAwned an;d maintaine -byEAL #2 The "site specific list of conmmunications methods" should include all communicationsmethods used to per-form inlitial emiergenc-y notifications to OROS as described in the site 59 EmnergencyPlan. The listing, shouild include installed plant equipment and compoeniits, and not items owned andmaintained by individuals. Example methods are ring down/dedicated telephone ines, commercialtelephone 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 emergencydeclaratio 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 usedto performF initial em:ergency' notifications to) the NRC= as desc;r-ribed in; the Site Emfergency Plan. T-he listing- shouldincluide inistalled plant equipmenit anld componients, anid not items owned and mainitained by individuals. These12' 1 4" "r o ; .. ....- ; T.:...... 41 --+/--: ,flXT~ ...---- ----,- *1 .+metheds me typteally the tietelephene lines,teated 1--meFgeney NEAH-teation ý-ý'stem ti-PiFsj teleplione kne and eemmeFeialECL Asinmn ~AatibLutes: 3.1 .1kC67 1 P a g e CA1ECL: lei4 ALERTInitiating Condition: Loss of (Feeter esel4RCS [PWR] OF RPV [BR)IA. .inventory.Operating Mode Applicability: CoAld ...td.n. Refioeling 5Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the Alei4-ALERT promptly upon determining that 15 minutes hasbeen 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 longerANDb. UNPLANNED ineFeaserise in (SITE SPECIFIC- SUAIP ANDWOR TANK) ANVANY of thefollowina 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 SumnpBasis:This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., aprecursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in thelevel of plant safety.EAL #1- A lowering of water level below (site specific level) elevation 32'- 9" indicates that operator actionshave 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 continuingdeerease reduction in water level will lead to core uncovery. Although related, EAL #1 is concerned with theloss 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 decayheat 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 byinstrumentation and/or power failures, or water level dropping below the range of available instrumentation. Ifwater level cannot be monitored, operators may determine that an inventory loss is occurring by observingchanges in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potentialsources 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 durationspecified in IC CSlIf the (reactor vessel/RCS [424W] or ,. PW tr]) inventory level continues to lower, then escalation to Site AreEmer~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 (+ 6inches above hot ley centerline). If RCS inventory is reduced below this level, normal decay heat removalsystems may not be available for core cooling. This threshold is not applicable to reduced inventory vacuum fillsince 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. Sinceother system leaks could raise levels in various tanks and sumps. the list was limited to the tanks and sumps thatwould have the highest probability of indicating RCS leakage inside the Reactor Containment BuildingAlthough 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 Leakage2. OPOP02-RH-0001, Rev. 63, Residual Heat Removal System OperationDeveloper- Notes:2Fer 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 standardoprainally significant setpoint at whichi some (typically high pr-essure ECCS) injection systems wouldaui-toma-Ptically stait and is a value significantly below the low RPY wvatei lev~el RPS actuiation setpointspecifed il IC CU 1..[......] The minimum all.wable level that supports opertian ofiinormally used decay heat removalsystems (e.g., Residual Heat Removal or- Shutdowxn Cooling). if multiple levels exist, specify, eac-h alongwith 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 movesthrouggh various operating modes-, -and- refuieling evolutions, pariciular-ly 9fo a PAIR. As appropriate to the plantdesigni, alter-nate means of deteiiiininig RCS level are inistalled to assure that the ability to MoNitor level withinthe range required by eper-ating procedures ;A411 noAt be interrulpted. The instrum~entatio rag ncssary to69 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.B70 1P a g e CA2ECL:Ae-t ALERTInitiating Condition: Loss of atIALL offsite and *tALL onsite AC power to emergency buses for 15minutes 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 hasbeen exceeded, or will likely be exceeded.(1) Loss of ALL offsite AND ALL onsite AC Power to (site specific emergency buses). oIALL three4160V 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 SYSTEMSrequiring electric power including those necessary for emergency core cooling, containment heatremoval/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-eaEmergcne.y STSITE AREA EMERGENCY because of the incFeasedadditional time available to restore anemergency bus to service. Additional time is available due to the reduced core decay heat load, and the lowertemperatures and pressures in various plant systems. Thus, when in these modes, this condition represents anactual 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 ICCSI or AS--RSI.CA2 -EAL-1 Selection Basis:N/AREFERENCES:1. OPOP04-AE-0001, Rev. 44. First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus2. OPOP04-AE-0004, Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus33. 0PSP03-EA-00,Rv '3). O-0002, Rev. 32, ESF Power Availability4. Drawing OOOOOEOAAAA. Rev. 24. Single Line Diagram. Main One Line Diagram. Unit No. I & 271 1Page Deveopes- Notes!F-or a power source that has mualtiple ggenertatos, the EAL and/or Basis section should reflect the minimunumfber' 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~edto feed 1 AC emergency bus), the EAL, and Basis section must speeify, that both gener-ators for- that sourcee ar-eoperatigThe "site spcfcemergency buses" are the buses fed by Off-Site Or em!ergency' AG pwer'e sourcees that supplypower I-- the electrical distribuition system tIhat power-s SAFETY SYSTEMAAS. There is typically 1 emer-gencyper 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 ofthis sourcee is controelled in acco-Ard-afce A-ith abno)rIml orF emerI-gency o~perating procedUreS. Or beyon)Id deSigH basisaccident 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 canbimplemenited 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 proeedurfalizedcapability to supply offsite AC power to an affected unit via a cro)ss tie to a companion unit may cr-edit thispoerAe source in the EAL provided that the planined cross tie strategy meets the requirem~ents of 10 C-FR 50.63.ECL. Ass~gignmentAtr~ibutes: 3.1 .2.BI721 P a g e CA3ECL:A~ei4 ALERTInitiating Condition: Inability to maintain the plant in cold shutdown.Operating Mode Applicability: Cold S.hutden, .. Refueling56Exa aple-Emergency Action Levels: (I or 2)Note: The Emergency Director should declare theA-lei4 ALERT promptly upon determining that the applicabletime has been exceeded, or will likely be exceeded.(1) UNPLANNED in RCS temperature to greater than (site specific Tec-nial Specifiatioen0-00ld shutdoWn., temlpe r.at limit) 200 0F (Tava) for greater than the duration specified in the-felewintab4e-Table C3.Table C3: RCS Heat-up Duration ThresholdsRCS Status Containment Closure Status Heat-up DurationIntact (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 isbeing reduced, the EAL is not applicable.(2) UNPLANNED RCS pressure inei-easerise greater than (site specifie pi....u.i. Frading)iLO si .(This EALdoes 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 RCSin excess of that which can currently be removed. Either condition represents an actual or potential substantialdegradation of the level of safety of the plant.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limitwhen 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 whenCONTAINMENT 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 thetemperature inefeaserise.The RCS Heat-up Duration Thresholds table also addresses an inefeaserise in RCS temperature with the RCSintact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS isproviding a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficienttime 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 reducedinventory [-P-z, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0minutes). This is because 1) the evaporated reactor coolant may be released directly into the Containmentatmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the topof irradiated fuel.EAL #2- provides a pressure-based indication of RCS heat-up.Escalation of theCS] or ARS.lAS-emergency classitication level EMERGENCY CLASSIFICATION LEVEL-RS]1.would be via ICCA3 -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 generatorU-tube draining, RCS level differences when operating at a mid-loop condition, decay heat removal systemdesign, and level instrumentation problems can lead to conditions where decay heat removal is lost and coreuncover can occur. NRC analyses show that there are sequences that can cause core uncovery in 15 to 20minutes, and severe core damage within an hour after decay heat removal is lost. The allowed time frames areconsistent with the guidance provided by Generic Letter 88-17 and believed to be conservative given that a lowpressure 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 determinedusing 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 accratelydctermined using installed intuctto.but not less than 10 sgFor PAWRs, this 1G wnd its associated EA~s address the concerns raised by Generic Letter 88 17, Less fTHewr Reiievf. ,A niumibcr ef phenefmena such as pr-eSSurization, voek&ing., steam gcncrater U tube draining, RCSlcaveal diffierfences w.hen oper-ating, at a mid leap condition, decay hecat removal system design, and leveinstrumcntation 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 coreedamage w~ithin an hour after decay heat r-em oval is lost. The allowed tiffc frames are cosistent with the74 1 P ag e guiidanee provYided b)y Genleric LetterContainmenit bairFier to fissieii produiECL Assignment Aittribates: 3.1.2.BT I II 1 1 ID ,I .I12 17 and behleve to b-e Consevtv ie that a IONS Sre--,.reet roicase is estaciisnee.75 P a a e CA6ECL: Alei- ALERTInitiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operatingmode.Operating Mode Applicability:Cold Shutdown, Refiieling 5,6Action 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 CoolingReservoir retaining dike along the North Wall" Other events with similar hazard characteristics as determined bythe Shift ManagerANDb. EITHER of the following:1. Event damage has caused indications of degraded performance in at least one train of aSAFETY SYSTEM needed for the current operating mode.OR4-2.The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structureneeded for the current operating mode.Basis:This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containingSAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces themargin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potentialsubstantial 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 itwill be readily available. The indications of degraded performance should be significant enough to cause concernregarding 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 readilyapparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators willmake this determination based on the totality of available event and damage report information. This is intendedto be a brief assessment not requiring lengthy analysis or quantification of the damage.Escalation of the emergeneyCSI orARSIASIRS1.classification le-'el EMERGENCY CLASSIFICATION LEVEL would be via ICCA6: EAL-I Selection Basis:The listed hazards are taken directly from NEI 99-01, Rev. 6. The only additional hazard was the inclusion of theMain 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 ProtectionDeve'loper Notes:For- (site _pecifi. hazar-d_). developOFSshould ecnsidcrf ineluding ather significant, site specific hazar-ds to theDulletee fist conipinco in LiL Ia ke.g., a setehe).N~eleaf PEAeFeplant SAFETY SYSTEMSeomprised of two or mor~e separate and r-edunfdant trains of.14 -AP-0-0-F-APP-3ea r-.+L-Q 1-31-46-26- W EleSlell eFlIeFia,LUkn, Lxz'sslonmnfii 'Lkrinates: -). I .-.77 1 P a g e CS1ECL: SiteArea Emergen.y SITE SITE AREA EMERGENCYInitiating Condition: Loss of (eaeteFvesselIRCS tPWR] or RPo,, [BWR]) inventory affecting core decayheatremoval capability.Operating Mode Applicability: Cold Shutdown, Ref-ieling 56E-4ample-Emergency Action Levels: (I or 2 or 3)Note: The Emergency Director should declare the Site Area Emergenty SITESITE AREA EMERGENCYpromptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.(1) a. CONTAINMENT CLOSURE not established.ANDb. (.Reaetai- vese RCS [DWR] or RPV [BIR] level less than (site specif. c level) 33% ofplenum.(2) a. CONTAINMENT CLOSURE established.ANDb. (Reae.taF ,essel,'RCS [PsWR] or- RP%1 [BWR]) l evel less than (site specific ievc!).(2) a. CONTAINMENT CLOSURE established.ANDb. RCS level less than 0% of plenum(2) a. 3) a. [PWR] or RPWN [Br3ARo) level cannot be monitored for 30 minutes orlonger.ANDb. Core uncovery is indicated by ANY of the following:,c......e specific r.adition ......) Reactor Containment Building. 68'-0" Area Radiation MonitorsRE-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 tanklevels in Table C2any, of the follo-ing OFnps or tank leNvels-of sufficient magnitude to indicate coreuncovery.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 SumpBasis:This IC addresses a significant and prolonged loss of (reactor vessel/RCS [PJM] Er RPM [BWR]) inventorycontrol and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCScomponent failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entailmajor 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 reactorcoolant boiling and a further reduction in reactor VesSel iev-eRCS level. If RCS/reactar vessel leveIRCS levelcannot be restored, fuel damage is probable.Outage/shutdown contingency plans typieal4y provide for re-establishing or verifying CONTAINMENTCLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specifiedRCSreacto: -'essel leveWRCS levels of EALs I.b and 2.b reflect the fact that with CONTAINMENT CLOSUREestablished, 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 abilityto monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions todetermine if core uncovery has actually occurred (i.e., to account for various accident progression andinstrumentation 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 instrumentationand/or power failures, or water level dropping below the range of available instrumentation. If water level camnotbe monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/ortank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow toensure 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 atCommercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guiclelines for Indust.y Actionsto Assess Shutdown Management.79 1 P a g e Escalation of the emergency classificaticn level EMERGENCY CLASSIFICATION LEVEL would be via ICCG 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 theRCS loop penetration at the reactor vessel. Six inches (6") below the bottom inside diameter of the RCS hot legnozzle (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 6inches above (Sensor 6) and 4.9 inches below (Sensor7) the prescribed elevation of 30'-6.5". When water level isat the desired elevation of 30'-6.5", Sensor 6 will be dry and Sensor 7 will be wet. This condition corresponds toa reading of 33% of plenum per OPOP02- 11-0002, RVWL Monitoring System, Addendum 1, RVWL SensorElevations.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 RCSlevel 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 MonitoringSystem thermocouple to TAF is Sensor 8 at elevation 29'-2.7". Use of RVWL to approximate TAF: with theinherent gap of 12 inches between indicated level and actual level, is acceptable for the purposes of signaling thatthe 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 arelocated on the 68'-0" elevation of the reactor containment building. Their locations are identified on drawing9C129A81105. Their range (0.1 mR/hr to 10,000 mR/hr) is identified in Table 12.3.4-1 of Section 12 of theUFSAR. 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 coreuncovery.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 fueland 6 R/hr at one foot above the top of active fuel. The high range of these monitors is 10 R/hr. The value of9.000 mR/hr was selected to ensure that the threshold is readily assessable and within the calibrated range of themonitor. The threshold value of 9,000 mR/hr corresponds to approximately 8 inches above the top of the activefuel with the reactor head on: which provides an additional indication that RCS levels are near the point of fueluncovery. These monitor readings in coniunction with the other threshold values allow for an accurateassessment of the EAL.Core uncovery can be determined by the secondary indications listed in this EAL. The secondary indicators ofinventory loss include a list of tanks/sumps found in OPOP04-RC-0003, Excessive RCS Leakage. Since other80 1P agPe system leaks could raise levels in various tanks and sumps, the list has been limited to the tanks and sumps thatwould 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 WaterLevels during Reftieling for EAL Thresholds2. OPOP03-ZG-0009, Rev. 59, Mid-Loop Operation, Addendum 1. RCS/RHR Simplified ElevationDiagram3. USFAR, Rev. 15, Chapter 12, Table 12.3.4-14. OPOP02-II-0002, Rev. 15, RVWL Monitoring System5. OPOP04-RC-0003, Rev 18, Excessive RCS Leakage6. 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 amouintof 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 variousoperating modes and refuleling ev'olutionS, pailicularly for A PA/R. AS appro~priate to the plant designIalter.nate .eans f determining RCS level are installed to asse that the ability to monitor level withinl therane rquiedby operating proEedureS Will not be interru~pted. Thke instrumentation range neessary--tosupport implementationi ofoper-ating, proceedur-es in the Cold Shutdown and Refuieling modes miay bedifferent (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" belothe bottom ID of the r-eactor vessel penetration and noet the low, point of the loop. If the availability of onscale level indication is suchi that this level value can be determined dur-ing some shutdowni moides Orconditions, but not others, thien specify, the mode dependent and/orF configuaration states dur~ingg which thelevel .Indcaio isf aplcbm h e ignad operation of water level inistrumenitationi is suchi that this lev'elvalue 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 onlscale 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 theleve inicaionis applicable. If the design and oper-ation of water level instrumentation is such that this levelv~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 willincrease. Enter- a "site specific. radHiation monitor" that could be used to) detect core unicovery and thieassociated ';site specific value" indicative of core uneovery. it is reeognized that the condition described bythis IC= may result in a r-adiation value beyond the oper-ating" orf display range of the installed radiatiomonitor. 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 thehighest ac ..rate monitorF readin... .This no0t-'ithstanding if the esti.ated,;aleulated monitorreading 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 bspecified in the EAL to build in an appro~priate level ofecorroboration between monitor r-eadings into theclsiiatin 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 fcrmaking such determination+,s.For EAL #3.b thir-d bullet Enter- any 'site specific sump and/ori tank" levels that couild be expected to chanigeif 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+.BAIRFo r SAL I-;. I.b "site specific level" is the Low Low Low EC-C- actuainson /Lvl1 h BAIR LowLEAS' LOW ECCS Aactuation- setpoint / Lev~el 1 was chosen becau'se it is a standard oper-ationally, significantsetpoint at which some (typieally low.N pr-essur~e EGG(S) inj~ection systems would automati cally, start andattempt to restor.e RP.M level. This is A R.PV w Ievel value that is observable below the Low, , evel 2value specified in IC CAl1, but significantly above the Top of Aetiv.e Fuel (TOAF) threshold specified iF=, 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~wewillinceas. Enter a "site spec-ific radiation monito4r" thatcolbeudtoeetcrenoeyadthassociated "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 radiationmonito. in those cases, EAL values should be determined with a margin sufficient to ensure that an accur.atemonitor r-ading is available. For example., an EAL monitor r.eading m.ight be set at 90%040 to 95(% o thehighest accur'ate moinitor reading. This proevision notwithstanding. if the estimated/lealeuilated moitrreading is greater than approximately, 1101% of the highest accurate monitor Feading. then developers maychoose not to include the monitor as an inidication and identify an alternate EAL thrfeshold.To further- pr-omote accrlate- Cl-ass'-ifispecified in the EAL to build ini o......... -n assessment., I I IVcat'Aon, leveloper'ssoudcnie it soee comblinationl ot montors could bean aeofeeiriate level Ctcroorto etwleen onorre-adi~ners into thePei- BWRs that do not bav.e installed radifatio n monitors capable 01 ind.i.atin core ..toeE)NeF', alternlate sitespeeitic 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 aretypically loeated below corFe miid planie, thiS May- net be a Viable inidicato' Of ecore un~e6very fo0r B3AsForF EAL #t3.b th~ird bullet Enter any "site Specific SUMP anidOr tanik" le'VelS that cOUld be exEpected to changeit- tfere wer~e a loss orf KFAr v mventeiry of sunil-eienm nagimay be included Of desir-ed.14HEle tO Hiwameae core uncovery. Speemc level valuesForf E-,4 #3.b fcurth1 bullet Deve!opers should determinie if other reliable indicator-s exiist to idenitify, fuielunever (.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 CG1ECL: Gener-al Emergency GENERAL EMERGENCYInitiating Condition: Loss of(reaeter- RCS [PW.] or inventory affecting ftiel cladintegrity with containment challenged.Operating Mode Applicability: C-ld Shutdown, Refueling;.5Exaompie-Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the General Em:crgeney GENERAL EMERGENCY- promptlyupon 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 30minutes or longer.ANDb. 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.ANDb. Core uncovery is indicated by ANY of the following:(site specific radiation moenitor) Reactor Containment Building. 68'-0" Area Radiation MonitorsRE-8055 or RE-8099 reading greater than 9.000 mR/hr.-s .p..ifi ae)..* Erratic source range monitor indicationOR4f-4W-UNPLANNED iner-easerise in (SITE SPECIFIC SUMP AND/OR TANK) ANY of thefollowing 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 SuraAND84 1 P a g e
- c. ANY indication friom Table Cl the C:ntainment Chall Tenge Table (see belb--W*.Table Cl: Containment Challene b-CONTAINMENT CLOSURE not establishedwithin 30 minuWtes,;* ** >(Explosive mixture) 4% hydrogen existsinside containment* UNPLANNED inefeaserise in containmentpressure-WSeeandary, eontainment radiation moanitorI % r r-eaoing above (site speeitie value) [l~IF CONTAINMENT CLOSURE is re-establishednrior to exceeding the 30-minute time limit. THENdeclaration 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, ,ta: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 withcontainment challenged. This condition represents actual or IMMINENT substantial core degradation or meltingwith potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAGexposure 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 reactorcoolant boiling and a further reduction in reactor vessel le-elRCS level. If RCS/reactor vessel levelRCS levelcannot be restored, fuel damage is probable.With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored releaseof 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 notrequired.85 1Pag e The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogenconcentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn willraise containment pressure and could result in collateral equipment damage leading to a loss of containmentintegrity. 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 couldresult in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service duringan event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gasconcentration reading as ambient conditions within the containment will preclude personnel access. Duringperiods when installed containment hydrogen gas monitors are out-of-service, operators may use the other liStedindications 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 abilityto monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions todetermine if core uncovery has actually occurred (i.e., to account for various accident progression andinstrumentation 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 instrumentationand/or power failures, or water level dropping below the range of available instrumentation. If water level cannotbe monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/ortank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow toensure 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 atCommercial Nuclear Power Plants in the United States; and N UMARC 91-06, Gutidelines /br Industrv Actionsto 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 RCSlevel 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 MonitoringSystem thermocouple to TAF is Sensor 8 at elevation 29'-2.7". Use of RVWL to approximate TAF; with theinherent Pgap of 12 inches between indicated level and actual level, is acceptable for the purposes of maintainingthe 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, ExcessiveRCS Leakagte. Since other system leaks could rise levels in various tanks and sumps. the list has been limited tothe tanks and sumps that would have the highest probability of indicating RCS leakage inside the ReactorContainment Building.As RCS level drops the dose rates above the core will rise. Area Radiation Monitors RE-8055 and RE-8099 arelocated on the 68'-0" elevation of the reactor containment building. Their locations are identified on drawing9C129A81105. Their range (0.1 mR/hr to 10.000 mR/hr) is identified in Table 12.3.4-1 of Section 12 of the86 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 coreuncovery.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 fueland 6 R/hr at one foot above the top of active fuel. The high range of these monitors is 10 R/hr. The value of9,000 mR/hr was selected for this threshold to ensure the threshold is readily assessable and within the calibratedrmnoe 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 thepoint of fuel uncovery. These monitor readings in coniunction with the other threshold values allow for anaccurate assessment of the EAL.REFERENCES:1. Calculation No. STPNOC013-CALC-006 Rev. 1. Dose Rate Evaluation of Reactor Vessel WaterLevels during Refieling for EAL Thresholds2. OPOP03-ZG-0009. Rev. 59, Mid-Loop Operations3. 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 Monitors5. OPOP05-EO-EO 10. Rev. 21. Loss of Reactor or Secondary Coolant6. OPOP04-RC-0003, Rev. 18. Excessive RCS LeakageDeveloper 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.edbyp perati..g prcedures will not be interrupted. The in,,trum..entati on range necessary to sppoi, implementationof operatinlg pro~edurfes in the Celd Shutdown and Refttelinig modes mlay' be diffeent (e.g., nialfewer) than thatr-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-nscale 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 iniainisapplicable. if the design and ope;ation of water- lev.el instrAmentation is such1 that this ...el v.alue canneo bedetermined at an .. time during, Cold Shutdown or Ref..eling m.des, then. do not include EAL .1 (classific.ationl 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 willincerease. Enter a "site specific- radiation monitor" that could be used to deteet corFe uncover-y and the associated871 Page retif; a aitinv lu'E bynd the operating, or display range of the installed radiation mon;itor. In thos)e caeSOSEAL 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 mtnitorr-eading. This provision n .twithstandi.g, if.the estimated/calculated monitor reading iS -Feater thanapproeximately 110% of the highest accur-ate monitor- reading, then developers may choose not to include themonitor 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 bwpecified in the EAL to build in naprrate level of corroboration between monitor- readings inito theclasifcatonassessment.For BIARs that do fiot have installed radiationi moinitor-s capable of indicating core uneoer~ey, alternfate sitespecific 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 formakng ucdeerinaios. BcueBRSreRage Monitor (SRN4) nuclear instruameantation; detectors arctypic-ally located below, core mid plane. this may not be a viable indicator of core uncovery forF BWAFor- EAL #2.b thir-d buillet Enter any, "site- specific sunip and/or tank" lev~els that could be exýpected to chiangeif there were a loss of inventory, Of suffic-ient magnitudle to indicate corFe uncover;y. Specific level values may beiincluded if desiredl.FEor EAL #12.b fourthi buillct Deve. peps shoulld determine ifothier realiab-lea indic-ators &eist to identify, fuelu1GNcover (e.g., remote viewn .sig.cmers) The goal is to idenitify any unique or site specific indications,noalr-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 aloss 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-ationinecessary5 to support a hydrogen burn (i.e., the lower deflagration limit). A concurr~ent containment exygenconcentration 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 releasethat my be indicative of a clhallenige to seondari.y containiment.EOP maximum safe values becauise these vausae fsl eeogThe "site specific value" should be based on thenizable and have a defined basis.& az I1
- 1 Alm4L- /kssianment Attrioutes: 4. !.4.88 1 P a g e 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)ICS/EALSTable E-1: Recognition Category "E" Initiating Condition MatrixUNUSUAL EVENTE-HU1 Damage to a loaded caskCONFINEMENT BOUNDARY.Op. Modes: .-4-ALL891 P a e E-HU1ECL: Ntificatin of Unusua! Event UNUSUAL EVENTInitiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARYOperating Mode Applicability: A41ALLEmergency Action Levels:(1) Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiationreading greater than: (2 times the site specific cask speeie.. tehnical specification allowable radiationlevel) on the surface cf the spent fAel cask. :a. 60 mnreni/hr (,ganna + neutron) on the top surface of the spent fuel caskORb. 600 mrern/hr (gamina + neutron) on the side surface of the spent fuel caskORb. 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 caskcontaining spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that theloaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to theenvironment, degradation of one or more fuel assemblies due to environmental factors, and configurationchanges 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 theTechnical Specification allowable radiation levels. The technical specification multiple of"2 times", which isalso used in Recognition Category RA IC AU-I-RU1, is used here to distinguish between non-emergency andemergency conditions. The emphasis for this classification is the degradation in the level of safety of the spentfuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extremedamage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based onmeasurement 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. Section5.3.2 of the "Certificate of Compliance No. 1032, Appendix A. Technical Specifications For The HI-STORMFW MPC Storage System", states:90 1 P ajz e 5.3.4 Notwithstanding the limits established in Section 5.3.3, the measured doserates on a loaded OVERPACK or TR4NSFER CASK shall not exceed thefollowiing values:a. 30 mnrem/hr (gamma + neutron) on the top of the OVERPACKb. 300 mnrem/hr (gammna + neutron) on the side of the OVERPACK,excluding inlet and outlet ductsc. 3500 mnrem/hr (waima + neutron) on the side of the CASKREFERENCES:1. Certificate of Compliance no. 1032, Appendix A, Technical Specifications For The NI-STORM FWMPC Storage System. Section 5.3, Radiation Protection Program. 10 CFR 72.104, Criteria ForRadioactive Materials In Effluents And Direct Radiation From An ISFSI or MRS91 1Page 9 FISSION PRODUCT BARRIER ICS/EALSTable 9-F-1: Recognition Category "F" Initiating Condition MatrixALERTFA1 A--yANY Loss or aiiyANY Potential Loss ofeither the Fuel Clad or RCS barrier.Op. AMfodes: P Opu6ati:i, Nat Sta:itf'y,tar-tmp. Hat4 S4Htdmiow 1,2,3,4SITE AREA EMERGENCYFSJ Loss or Potential Loss of anyANY two barriers.Op. AMfodes: Poi'e"r Oper'atio, ot Sta& by,______ S ..tt up, f tJS, ,. ....w... 1,2,3,4GENERAL EMERGENCYFG1 Loss of anyANY two barriers and Loss orPotential Loss of the third barrier.Op. Modes: Power Oper'atio.'a, ......S.. 'n at Shue,.wn. 1,2,3.4See Tabic 9 F 2 for- BMIR EALsSee Tanble 9 F 3 femr PMIR EALsDev'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 aidto facilitate timely and accurlate classification of fission productbarrifr losses an/3ptnillsses. Suchl aids are typically,comprised of logic fie'w, diagrams. "scoring-". rit',ria or heekboxtyýpe mnatricies. The user- aid logic must be cositetwih ht of the.aEaeent diagr-am.92 1 P auý e De~velpep Notes1. 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 ContainmentBRFr-e4-."Unusual Event ICs associated with fission pr-oduct barriers are addressed ini ReconiioCaegery S.2.- For aecident conditions involving a radiological release, evaluation of the fission proeduct barrFierthresholds will need to be per-formed in cojnto ihdose assessments to ensur~e corr-ect andtimely escalation of the emiergency classification. For- example, an evalbuation of the fission productbarrier thresholds may result in a Site Area Emergency classification while a dose a sse ssment may0indieate 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 specificdesign and operating- characteristics. This may r-equir~e that deveJeper-s reate different thresholdsthan 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 combinationsand classification outcomes sho.n.. in t ...B.R or PWR EAL fission product barrier tables. TheNRC staff considers the presenitation miethod o~f the Recognition Categorfy F information to be animpo ..ant use- aid and may request a change to a parti;.lar. proposed .method if, among otherresns, 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 typesdefined in Technical Specifications but also includes the loss of RCS mass to any) location insidecontainm...ent, a secondary side system (i.e.. PWR steam. .generator tube leakage), an inter.facinsystem, or outside of conainment. The release f liquid oer steam mass from the RCS due to the asdesigned...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 ohowfarR pesentcnditions Are from meeting a threshold that wouild require a General Emfergenceydeclar-ation. For exNample, if he Fuel Cl.a-d -And- RCS fissioni product barriers were both lost, then thereshould be frequtent assessmienits of conltainment radioac.tie inventory and inert.Alternatively, ifboth 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 conditionsshould be fmaiiatained. For example, a steady incere-ase in RCS leakage would rpentan. icr-easmnvrisk 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 EMERGENCYA n' , Les oF an Potential Les.7 .7a. Loss of PotentiaeiteF he ietGla r RG bamrirI LOSS Ofall)'b. Less of any two barriers -anLess or Potential Less of theth*ld ba!--e!..F~iA CadBiiric dRCSýaRricr Coini nt Biarriei*f .OS POU-EN h-. LOSS i -POTEN LOSS POTENTIALTIALTIALLOSSLOSS LOSS4.P-~P*afI-YConta~ifnment Radiationt j.---4.-Pr-*fima*Containmcnt Radiation k. 4.P-Pifffl~i, Containment RadiationA Pr;imar' ...ntainment 1. .. A. P,,imary ..ntainment ill Not f N A Pr..i.ay +ontainmentradiatiOn m110ito01 Applieab radiatiOnl MOitor A-ppli-ea App~li~eab4 radiation 010nit40rreading 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, asITF M7applieaeyeable a~- p~a~)p eabt4i-. 6.... E Di..e.tor- judgment s. 6. Emcrgeney Director judgment t. 6. Emcr- ,y Drector JudgmentA. 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 inthee oiin afthe opinonefflie Oea-flil ffi he oinion a OeifEMefthe til~eopiniEiftl Ofgmerg-@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 niaeof the Fuel Cla.Potential Loss ofthe the RCS B... Potential Loss of the the Co..ainment Potential Loss cf.thegafi-ef -Fuel Clad Barrier. RC+Ii.e -3A-FleiA -CItntainment Barrier.f-,94 ....P.a ...
Basis information ForBWIR EAL Fission Product Barrier- Table 9 F 2BMIR 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 ActivityThis thireshold indi2ates that R. S radioacgtivity is greaer than 300 '-Ci/gm dsse equivalent " 13-1. Reaetorcoolant activity above this level is 'greater than that expected for- iodine spikies and corresponds to an approimeinate r-angc of290 to 51% ftuie clad damiage. Sinee this condition indicates that a significant amount of fuel clad damage has occur-red, itrepresents 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 I13 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 radiatiomnitor readinig componenit.~Add this paragraph (orF similar v'crE)ding-) to) the Basis if the thr~eshold includes a sample analysis component, "It isFrecognpizied that sample collection and analysis of r-eactor- coolant with highly elevated activity levels, could requir-e sever-alhours to comsplete. Nonietheless, a sample related thrfeshold is included as a backuip to other indications."2. RPV Water LevelThe Loss thrfeshold i'cpresents the EOP requirement for pray cotiment flooding. This is identified in the BsAROGEPGsiSAGs 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 the90P need to flood the primar cot imet when RPW water level cannot be determinied and core damage d&ie toinadequate core cooling is believe-dA to bePotential Loss 2.AThis 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 PotentialLoss of the Fuel Clad barr-ier anid a Loss of the RCS barrier that appr'opriately' escalates the emer-gencey classification levelto 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 restoredand maintained above the specified level following depressurizatio~n Of the RPW (either- manually. auitomatically or byfailr~e of the RCS bairier) or when procedural guidance or- a lack of.low pressrUre .RPM. ii.e.ti.n s.ur. pi ..luEmergency R42 dep.. ess.ization. EOPs allow.. the operator A w..ide.choice.of... in.c ..tion sources to consider wheni-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 preventrPestoation of RPe waterl ievel until pressure drops below the shutoff heads of.available iection. souces. Ther.efo.e, thisFuel Clad barrier Potential Loss is met only after- either:ý I) the RPMI has been depr-essur-ized. or- requir-ed emer-gency RPMld4.....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 RPMdepressurization 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 abovNethe specified limit (to~p of active fuiel). The dtriaoneursanevaluation of system per-formance and availability inrelationi to the RPM water level value and trenld. A thr-eshold prescribing declaration when a thrfeshold value eaiunot berestor-ed and maintained above a specified limit does not i-eqair-e imfmediate actioni simply because the currFent valuie isbelow the top of activ~e fuel, but does not pei-mit exEtended oper-ation below the limfit; the thr-eshold muist be conHsideredreachied 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 thletop 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 theFuel 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~illdictate 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 fuelclad barrier, a potential less of the feel clad barrFier is specified.Dev-eloper Notes:!Loss 2.AThe phrfase, "Prifimary containment flooding requrired,"' should be modified to agree with the site specific EOP phrsexit fromf.. all .OPS and ent ry4to the SAGs (e.g., drywell flooding r-equir-ted, etc.).Potential Los 2 AThe decision that "RP3. water level cannot be. detearmined" is directed by guidance given in the RPW3 water- level controljctins of f the e PBsi.3. Not Applicable (included forF numbering consistency between barrier- tables)41. Pr-imar~y Continmienit RadiationThe 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 %fecld dam~age. Since this condition indicates that a signiificant amouint of fue la amg has occurrFed, it represents a lossof 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 itinidicates a loss of both the Fuel Clad Barrier and the RCS Barrier-.Note that a combination of the two monitor-radngappropriately, esc.alates the emergelncy. classification level to. a Site Area Emergency.T-here is no Potential Loss threshold assoeiatcd w.ith Primiarfy Containment RadiatiefonDeveloper Notes:The r-eading shou~ld be determnined assuming the instantanieous r-elease and dispersal of the reactor eoolant noble gas andiodine inv.nteiy, with RCS radioactivity concentration equal to 300 ..i.,/,m dose equivalent I 13 1, into the primarycontainment atmosphere.5. Other indicationsLoss and/or- Potential Loss 5-.AThis subeateggory addresses other site specific thr-esholds that may be included to indicate less or- potential loss of the FuelClad barr-ier- based on4 plant specific design chiaracteristics not considered in the generic gui danc-e.Developer Notes:Loss and/or Potential Loss 5-.ADe~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:~uiuof 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 thiscoIb:lumn Basis informationi for the other thr-esholds may be used to gauge the relati've bairrier thrfeat level.6. Emerg-ency Director juidgmentLess-6.AThis thrfeshold addresses any, ethcr facetors that are to be used by the Emfergeney Dir-ector in determiininig whether the FuelClad Barr-ier is lost.Potential Loass 6.AThis thrfeshold addresses any, other facetors that may be uised by the Emer-gency Dir-ector in determinifing whether thle FuelClad Barrier is otnilyost. The Emergency Director- should also consider whethier or not to declare the bat:rierpotntillylos 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 systempiping ":p to and including the is-lation valves.1. Primary Containment PressureThe (site specific value) primar:y ..ntainment pr.essure iS !the dry', well high pr.essure setpoint Which indicates a LOCA byautomatica.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 LevelThis 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 thresholdindicates a Loss of the RCS barrier and Potential Loess of the Fuel Clad barrier and that appropriately escalatesthemcgeney 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.edand maintained above the specified level followinhg depressurization of the RP, (either mafnu.ally, automatically oF,-"bfailure of the RCS barrFier) Or when pro)cedural gulidance or a lack of low pressurfe RPMP injection sources precluideEmergeney RPM depressurization EOPs allow the operator a wide of RPM injeetion sources to consider whenrestoring RPM water level to within pr.eseribed limflits. EOPs also specify depressut.ization of the RPM in order to facilitateRP ... .at. level -it "+'-. pressure i-jection som-rces. some events, elevated RPM pressure may preventretoationi of RPM water- level uintil pr-essur-e drops belowN the shutoffheads of available injectioni soureS ThefetrBf. tRCS barrier Loss is me" o.n, _a4fe either:i 1) the RPMo has been depressu.ized. or required emergency RPMdepfessurization has been attempted, giving the operateor an ,pp..unit. to ass" ss the capability of 1' pressure ijectionsou.ces to resto. e RPM, w,.ater. level -or 2) low pressure RPM' iýection systems are available, precuding, RPDdep.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 abovthe specified limit (top of active fuel). The determinatio' requires an aluation ofsystemn perfomane and availability inrelation to the RPM w.ater lvlvalue anid trenid. A threshold prescribing- declaration when a threshold value eaonno berestered and maintained above a specified limit does not require immediate action simply because the current value isbelew the top of activ~e fuiel, but does noet permit etended opei-ation beyonid the lkimit; the threshold must be consideredreached 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 thetop of active fuiel in orFder to r-educe reactor power. RPM water- level is then contro~lled betwe~en the top of active futel andthe Mnimutm Steam Coolin-g: RP. Water Level (MSCRAL). Although such action is a challenge to core coorin-g an-d theFulel Clad barrier, the immediate need to r-educe reactor power, is the higher prk)ior'. For suchl events, W~s SAS orF SSS willdictate the need for emergency classifieation.There is no RCS Potential Loss threshold associated with RP,, Water Lev.el.RCS Leak RateLoss Threshold .A,Large high;I energy lines that ruptuire outside primary).' cotainm.ent clan dischar.ge significant amounts of inv.entoery ajeopardize the pressure retaining capabilit,, of the RCS until they ar-e isolated. if it is determt~ined that the ruptured linecannot be p-ofptly isolated fom. the Contro;.+el Room, the RS ba r..ri. Loss threshold is et.Loss Thrieshold 9.BEmergency RPV Defressurization in accordance with the EOPs is i.dicative of a loss Of the RCS barrier. if EmergeoncyRPM 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 thediminished 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 -EOPtem;peratrfe. o~r raito a omlOperating values in areas swell as main; steam line tunnel, RCIC, MPG!, oet.. whichindiee~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 plantoperating 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 Warranitanl 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*nun~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.hencoemb-inedd v.'ith CoAnt-Aimmeant Barriei- Loss thre-shold- 3.A4 (after a coentainment iso-lation) and a Gener-al Emer-gency whenthe 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, canrapidydpesrz the-a R W.. These- lIne are typically isolated by actuation of the Leak- Detection systmLarge 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 areUNISOLABLE represent a signifkicat los f the- RCS-& barrier-.4. Primary' Containment RadiationThe r-adiation monnitor readinig corresponds to an inistantaneous release of all reactor coolant mass into the primaryLcontainment, assuming that reactor. coolant activity equals Technical Specification allowable limits. This value is lowerthan 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-imarCntDcvclopcr Notcs:The r-eading shouild be determ~ined assumning, the instantaneouls release and dispersal of the. re-actor coolant noble gas anldiodinie 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'athis 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' besuch that r-adi-ation from -A cloud- of released RCS gases cannot be distingu~iShedp- from- rad-i-ationj emanajjtinjg from piping andcomponents 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 IndicationisLoss and/or Potential Loss 5.AThis subcnte-geiy addresses other- site specific thrfesholds that may' be included to inidicate loss orF potenitial loss of the RCSbarrier- based on plant specific design characteristics not consider-ed iii the generici guidance.Developer- Notes:2Loss and'or Potential Loss 5-ADevelopers 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 uniuOr 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 thiscolumn. Basis iifomiation forý the other- thr-esholds may' be used to gauge the relative barrfier threat lvl6. Emerggency Dir-ector judgmentLoss 6.A98 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 RCSbarrier iS 10st.Loss 6.AThis threshold addr*esses any) .ther fat .rs that may be used by the Emer.gen. y Di-e.tor in detemining whethe. the RCSBarrier is p"tentially lost. The Eme..gen.y Dire.tcr- should also c.nsider ...ether 0!r not to deelare the barrier potentiallylost in the event that barrier; stat. s .annot be m.nitor.ed.Developer Notes:NetiThe Primary' ContAi;nmen Bar..ier inl.,des the diywel. , the wetwell, their respetive i...r.paths, and otherconnecetions up to and including the outermost containment isolation valves. Containment Barr-ier thresholds ar-e uisedaeriter'ia for- escalation of the E,-- fi..m Alek to a Site Area Emergency or a Gen.eal Emergency.1.Pr-imnary Containiment ConditionsLess 66t and 1ABRapid 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 pressureshould increase as a result of mass and e..ergy.. r elease into the primar-y .ontainment from a LOCA. ThRus, primar.yecntainment 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 inot assigned. The une.pected (UNPLANNED) r.esponse is impoi0ant because it is the indica;.tor for .a. tainment bypasseanditien.Potential Loss ] AThe thr.eshold pressure is the primary e.ntain.men.t in.teral design pressure. Structur.al aeeptane. testing demonstrates. thec..pability of the primary contain.ent to resist presslres greater than the inte..al design pressure. A pressure of thisand iseter than those expected to result from ay design basis aciden. t and, thus, represen.t a Potential Loss othe 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 richenvionmet, apotenitially exiplosiv~e m:ixture exists. if the com~bustible mixEture ignie inietepiaycnanetv~ ~ ~ ~ ~ ~ ~ ~~~~~~~~~~i .................................. ies inside the primary containinent,loss of the Containment barrie;" could occur.Potential Loss I .CThe Heat Capacity Temperature Lim.it (l-IT. ) is the highest s.pp.ession pol temper.atue fo.m which Em.er.gen.y RPDepressurization will not raise:a Suppression chamber temperature above the maximum temperature eapability of the su.ppr-ession. cham..ber. andeipe w....ithin the su.ppression chamber which may be required to ope.ate when the R... is pressurized,OR0 Suppr.ession chamber pressre above Prima.r.y Co.tain.ent Pressue Limit A, while the rate of eneg. transfer frmthe 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 topreclude failure offthe containment and equipment in the containment necessary for the safe shutdowin of the planit anther-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-ationsof.hydrogen and o....gen.. FOr ,1 I/W I .. ntainmen.ts the deflagration limits are "69; hydrogen. and 50/o oxygen in tdywell 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 !.CSincee the HCTL6 is defined asuig age of suppression pool water levels as low as the elevation of the downeemefoffenings in MLE 141 containiments, or- 2 feet above the elevation of the horizontal v~ents in a N&M H! cntakinment, it isunnecessary to consider separate Containment barrier Loss or Potential Less thr'esholds for abnormal suppression pool99 1 P a g e water- level conditions. if desired, developers m~a:, incleude a separate Containmenit Potential LOSS thrfeShOld baSed an theinability to mnaintain suppression peal wvater level above the dov.ncomer openings in 4.1- 1111cotinens orf 2- fee~t aboveathe elevation Of the horFiZontal venAts in -A MLIll W ontainmient with RPNV pr-eSSure abovNe the inmmdecay heat r-emovalpressur'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'-2AThe Potential Loss thr-eshold is idenitical to tile Fulel Clad Loss RPYl Water Level thr~eshold 2.A. The Potential Los1-qi-i tfor Primar-y C01anetFlooding~ ind1cioaes dqaeCa- al cannot be restored and maintainled anldthat core damage is possible. BAI'R EPGs/,kGSA sp~ecify, the conditions that r-equir-e primary containment flooding. Whenprima~' containment flooding is required, the EPGs are e-xited and SAGs are entered. Entry into SAGs is a logialescalation ini response to) the iniability to r-estore anld maintaini adeqtiate corFe ceoo 1g. PRA Studies indicate that theconditioni of this Potential Loss threshold eould be a cor~e mfelt sequence whichi, if not correceted, could lead to R-PN failureand increased potential forpimar onanment failure. in conjunction with the RPV water level Loss thresholds in theFuel Clad and Cs barierclms this, threshold resuilts in the declaration of a Genieral EmergecyDevlper-NotesThe phr-ase, "Primary conitainment floodinig r-equired," should be modified to agr~ee with tile site specific EOP phr~aseinidicating e~xit fr-om all EOPs and entry to tile SAGS (egdy Pelfooding- required, etc.).~.Pri-iar,' Containmient Isolation FailureThes theshldsaddress incomiplete containment isolation that allows an UNISOLABLE direct release to theThe use of the moditier "direct" iln defining the release path discriminates against release paths through interfacing, liquidmOFino-elease pathways, sul siit-iiftlines, Htprotected by thePifHFC-Htilel !eafi ytfiThe 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., retentionability 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 bm~inrE radielegieal releases associated with allowable primary) containment leakage through various pefietrations or systemfcompoenelts. Minorf releases may also occur- if a pi-iimar:, containment isolatiain valve(s) fails to close but thie pr-im~arycontainment 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 radioactivityr-elease rate limits will be exceeded. 914del thes@e onditions With a Valid primary) containmentOft i-sol1ationl Signal, thlecotainmenit 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 secondarycontainmnent and/or thle env~ironentlf.NA is -A Loss of the Containment. Venting for- pr-imary containment pressure conitrol whennot inl an accident situation (e.g., to control1 pressure below the dryv.'~ell high pressur-e scr-am setpeinlt) does not mieet thethresheld condition.The Max Safe Oper-ating Teprtr n h a ae Operatinig R-adiation; Level Ave each; the higst value of tesp3aram~eters at whichi neither: (1) equipmient niecessary for the safe shubtdown of the plant will fail, nor (2) personniel accessIleeessar-y for- the safe shutdo'.n of the plant, will bepreluded. 9OAs utilize these temperatures and radiation4 levelstestablshcod-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. Aprimary system is defned to be the pipes, valves, aid other- equipment whichi connect directly to the RPV such thaar-eduetiocn in PLPN pressur-e will effect a decrease i-n thteam or water beking dischar-ged thrfough anl unisolated breakE in tleeomination 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 LesDeveloper Notes:Less 3.B4rhes lie 1,Aassociated with Primai:' ..ntaimn.nt isolation FailureConsideration may be given to specifyinig the specific procedur-al step within the Primar-y Contaiimnment Control BOP thatdefines intentional venting of the Primar-y Containment reegardlcss of off-site r-adioactiv~ity release Fate_.4. Primary Cont.ain.ment RadiationThkere is no Less thireshold associated with Primary Containment Radiation.Potential Loss '.AThe r-adiation monitor- readinig corresponds to ani instantaneous release of all reactor coolant mass into the primaryLcontainment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is .e.. abov tha. used todetermine 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 fuelclad failure mu.st be greater than approxim.ately. 20%.; in order for there to be a mjo.. r r.elease of radioactivity .equ.iringoffsite pr-otective actions. For this condition to exist, there must alreadY' have been a less of the RCS Barrier and the FuelClad Bar.ier. it is therefore prudent to treat this condition as a potential loss of containment which would then escalate theemergency' 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 tilebasis for- usinig the 200% fuiel claddinig faýilure value. Unless ther-e is a site specific analy'sis juistifyinig a different v~alue, thereading should be deter-mined assuming, the instantaneous release and dlispersal of the reactor coolant noble gas and iodininv.entefry associated with 2-0% fuiel clad failure into thlpimay contaMinment atmosphere.S. Other indicationsLoss and/0r Potential Loss 5.AThis su.beategor' addresses other site specific thresholds that may be included to indicate loss or potential loss of theContainment bareier based on plant spec.ific d.si characteristics not considered in the generic guidance.Developer Notes:Loass ado Potential Loss &.ADeveloper-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!niuosite 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 thiscolumn. Basis information forf the other thresholds may be used to gauge the relative barrfier- threat level.6. E m ergency' Director- JudgmenitLess-6.AThis threshold addresses any otIher facetors that are to be used by' the Emergency' Director- in determining whether theContainment barr-ier is lostoPotential Less-".This threshold addresses any' o ther factors that may' be uised by' the Emerggency' Direector in deter-mining whether thContainmienit Baresptnillyelst. The Emergency' Diretor should also consider whethier or not to declare the barrierpotenitially' lost in the event that barrier status cannot be monitor-ed.Developer- Notes:Noee101 Page Table 9-F-23: p3AWR EAL Fission Product Barrier TableThresholds for LOSS or POTENTIAL LOSS of BarriersFA1 ALERT FS1 SITE AREA EMERGENCY FG1 GENERAL EMERGENCYA-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 orthe 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 LOSS1. RCS or SG Tube Leakage 1. RCS or SG Tube Leakage 1. RCS or SG Tube LeakageNot Applicable A. ,CS/reactor- ve*,,se A. An automatic or A. Operation of a standby A. A leaking or Not Applicablelevel less than (site manual ECCS (SI) charging (makeuttp RUPTURED SGCore actuation is required pump is required by is FAULTEDCooling -Orange ent ry by EITHER of the EITHER of the outside ofconditions met following: following: containment.1_. UNISOLABLE 1. UNISOLABLERCS leakage RCS leakageVT OR OR4-.2. SG tube 2. SG tube leakage.RUPTURE. ORaB. R.CS cocedew'ni'ate greater than (siteSpecifiC prCeSuriZQelthlei-mal shockidieatieo-ns). Integgity-Red entry conditionsmet121 ~P age .E4 Fuel Clad Barrier. RCS Barrier (64 Containment Barrier(44) LOSS (4) POTENTIAL , LOSS (K4) POTENTIAL (-) LOSS (M) POTENTIALLOSS LOSS LOSS(N) 2. Inadequate Heat Removal (-) 2. Inadequate Heat Removal (-P) 2. Inadequate Heat RemovalA. @** ei A. C Not Applicable A Not Applicable A.. .spe..,feang&l+ : i-ei -aef eap-4iryhv-4ean into co..e -.olingtemper-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 .notconditions met entry conditions efFective within 1-OR met. minutes. CoreB. i Cooling -Red entryheat re.!lo-al conditions met forheapa t v15 minutes orvia ctcamlog.I eat Sink -Redentry conditions met3. RCS Activity / Containment Radiation Fuel 3. RCS Activity / Containment Radiation RCS 3. RCS Activity! Containment RadiationC-,d 'Ba2iefr BF*e B- ieeAl. C Not Applicable A. Ce,+ti+-n'e'+4t Not Applicable Not Applicable Al. Gontaii'in,tr-adiation 1*monito faiatio)i' infltont.. radiationi mnitorredfi ae,!eadina 1greater i-eadin -eate~rt h m ( si t s ,. , 0 ......fl a.,(:eLpe ie-H--e+.-RCB Rad value). RCB R-a- a-e+l.RCB RadMonitor RT-8050 oi;itoi. RT 84050 Monitor RT-8050or RT-8051 efjTgýM or RT-8051greater than- 40 .... +la 2 .gýreater than 380R/hr R-*Not R/hrOR Applicable OR2. HATCH 2. HATCHMONITOR MONITORgreater than 3-00 greater than 4-20m R/hr 840 mR/lhr1031 P age ORindieatlvs idates4-34, Sampleanalvsis indicatesthat reactor coolantactivity is ireaterthan 300 ItCi/gmdose equivalent I-131.104LPage S Fuel Clad Barrier : RCS Barrier Containment Barrier,LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS4. Containment Integrit or Bypass 4. Containment Integrity or Bypass 4. Containment Integrity or BypassNot Applicable Not Applicable Not Applicable Not Applicable A. Containment A. Centainmietisolation is required pi.es.... greater thanAND EITHER of (ste .p.cii. ;ev.l,,the following: Containment -Redentry conditions metI. Containment ORintegrity hasbeen lost based B. Explosive mixtureon Emergency exists insideDirector containmentjudgment. (H,> 4%)OR ORC1. Containment2. UNISOLABLE pressure greaterpathway from than 9.5 psig. (sitethe containment speei... pressto the setpoint)environment ANDexists.2. Less than one fullOR train of(&ieB. Indications of RCS .. .e ..... .leakage outside of e... ....containment. Containment Sprayis operating perdesign for 15minutes or longer.105 1 P ag e Fuel Clad Ba.rier RCS Barrier Containment LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS5. Other Indications 5. Other Indications 5. Other IndicationsA. 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/A6. Emergency Director Judgment 6. Emergency Director Judgment 6. Emergency Director JudgmentA. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition inthe opinion of the the opinion of the the opinion of the the opinion of the the opinion of the the opinion of theEmergency Director Emergency Director Emergency Director Emergency Director Emergency Director Emergency Directorthat indicates Loss that indicates that indicates Loss that indicates that indicates Loss that indicatesof the Fuel Clad Potential Loss of the of the RCS Barrier. Potential Loss of the of the Containment Potential Loss of theBarrier. Fuel Clad Barrier. RCS Barrier. Barrier. ContainmentBarrier.106 I P a R e Basis Information ForPWR-EAL Fission Product Barrier Table 9-F-23Devecloper- Notes:Threshold Parameters and ValuesEa.h WA.R owner's gir...p has developed a mctcdeloy f, guiding the developm.ent ad impleientationof Os(i.,aesigpatpaetrand determiining and prieritizing oaperateF ac~tions). Many of thethresholds coitained iii the PWR EAL Fission Product Bar'rier Table rFelect conditions that arespec-iflcally Addressed6 in EP__12 (e.,g., a less, of heat r-emoval eapability by the steamf gener'atorS). Whendeveloping a site specific threshold, developer-s shouild use thle parameter-S and values specified withinltheir- EONs that alignl wit th1 cOW dto Ele::;ibed byý the generici threshold anid bas-is. adR1 latedevelaper notes. This approachl will enueensistenc btwe the site specific PONs and enierzeneyC]',assifi-C',tion schleme. 'and' thuIs 4teci[tat'e timeb,) andJ aeeiiEc e::atccsisifieatioti assessnieflt'sem-ag-htemntatontheWestinghoeie- Arup(-)G) develepe4a 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-apl-oritized anid systematic mianner. 'Hie M106 Critical Safe~ty Fuinetions al-c prFesentedl beON'.SSubce'itica!it:,'-Cre" Coo!inkRCS integrity'Con1tainmlent-RCS inventeryThe WOG- ERGs provide a methodology forf monitor-ing the status of the Cr-itical Safety Fu1nctions aridelassit'ilig thie signifi.an.e of a challen.ge to a f.netio... this, is r.efer.ed toe CriticalSaf~ety Fu4nction Status Trees (C-SFSTs). Peo- plants that have imiplemented the WOG E=RGs, thle guidancei,901 allows for. use a CSFST assessm..ent ,esu.lts as EALs. and fissbion product barrfSeiel;ess/potentia 1O !H thi, Manner. an eer-gen,'y classifie.ation assessment miay fllodictlyfrm a CSFST assessm'ent.It is iMpor'tant to unlderStafrd that tile CSFSTs are-e ealuated uising plant par'aieter-s, andthat they arf-pl-vend pe for puraposes of drivinge....g.... .operating .......... usage. F _-. .... ..e.cnditions of inter.est, thle ..e..i tr1,esholdswithi' 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, aC'F9ST ter'Imlinus m.ay lbe used as well. Por this -easn, inclusion. of thle C ..ST related thr..eshold wouldbe 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 potentialloss 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 WOGCSFRST 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 conditionsE)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.. 2ti.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 -vathre..olds muti,1 be...... rsd in..... the table.i (e.., it is ...: permissible to use nl the C .1 t s as a107 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 thattI-Fe4ietlo,-,; t!STP iGuidSincebreshOld 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 Responseelines (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 CSFSTassessment results as EALs and fission product barrier loss/potential loss thresholds. This approachallows consistency betwe en OPs and emergenvc classifications.108 P a e -I-R FUEL CLAD BARRIER THRESHOLDSThe Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.I. RCS or SG Tube LeakageLoss 1There is no Loss threshold associated with RCS or SG Tube Leakage.Potential Loss I.Ain 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.A41. Enterz the site specific r'eactor vessel wateir level valuEe(5) uised by E ON, to identify, a degr-aded corecooling condition (e.g., 'equires prompt reSteration action). The rcactor vessel lel that correspandto 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 .pnthe status of RCP~s, if applicable').6. Westinghouse ERGC Plants,7.2. Inadequate Heat RemovalLoss 2.AThis reading indicates temperatures '.vithini tke core Core Cooling -Red en1200" F) are sufficient to cause significant superheating of reactor coolant.try conditions me-t-CEl's >Potential Loss 2.AThis reading indicates teiperatures Within tile Core Cooling -CFjare sufficient to allow the onset of heat-induced cladding damage.)range entry conditions (E.Ts > 708'Potential Loss 2.BHeat 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 challengeto the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heatsink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, theremay be unusual accident conditions during which operators intentionally reduce the heat removalcapability 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 thisthreshold is identical to RCS Barrier Potential Loss threshold 2.A; both will be met. This conditionwarrants a Site Area Emergency SITE AREA EMERGENCY declaration because inadequate RCS heat--109 Page PWR FUEL CLAD BARRIER THRESHOLDSremoval may result in fuel heat-up sufficient to damage the cladding and ieieaeraise RCS pressure tothe 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 onaefingthe number or .therrt.fe.p .radig aeeesary tO drive aci;o... (;.g.. , readinggr.eater than 1 .200912 is required ber... g to an. inadequate c..re cooling procedure). Tomaintain eensisteney with EOPs', these decisioni irnaking 4riteria may be utsed in the. eeare exittherm,..c.,.uple :'eading thresholds.La cc 2. AEnter a site speeific temper.atre value tat corr.esponds to signi..can. ill 00ore SUpe.eatilg of .eactogcoolant. may also" be used.For plants that hav..e implemented Westinghouse Oners Group Emergenmy Response Guidelines, enterthe parameters and values used in the Core Cooling Red Path.Potential Loss 2.AEnter a site specific temperature value that corresponds to cre. conditions at the on.set. of heat inducedcladding da.. age (e.g.2hmru -hal fortespodsn foreation ofsuperheated steair assuming thatheRCS is intact). 7004' may al.so be used.For plants that have ime 1,mnted etihg house O wner.s ..oup Emergency Response GCidelines. enterthe parameter.s and values u.sed in the Cor.e Cooling. Oran.ge Path.Potential Loss 2.REnter the site speciic and values that define an. extr.em..e challen.ge to the ability to removeheat fi-om the RCS via thle steam generators. These will typically' be parameter-s and v, alues thai~t wouldrequir-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 Plants6s a os inldication,. developerg shouild conisider- inceludinig a thr-eshold the. same as, or- similar- to. '-CoreCoaling 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 RadiationLoss 3.A. 1The radiation readings for the containment high ranse area monitors (RT-8050 and RT-805 1) correspondmonitor 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 activityabove this level is greater than that expected for iodine spikes and corresponds to an approximate range of2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damagehas occurred, it represents a loss of the Fuel Clad Barrier. The values for RT-8050 and RT-8051 werebased on Calculation STPNOCO 13-004 Rev.2. The threshold values were conservatively rounded within2% 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 radiationdose 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 THRESHOLDSLoss 3.A.2The HATCH MONITOR is located outside containment and is the back-LIp monitor to the containmenthigh range monitors (RT-8050 and RT-805 I). The HATCH MONITOR threshold value is based onCalculation No. 03-ZE-003. This value corresponds to the calculated containment high range monitorreadin.zs for Fuel Clad Barrier Loss 3.AThe radiation monitor reading in this threshold is higher than that specified for RCS Barrier Lossthreshold 3.A since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that acombination of the two monitor readings appropriately escalates the emergency classification levelEMERGENCY CLASSIFICATION LEVEL to a Site Area Emergency SITE AREA EMERGENCY.Loss 3.BThis threshold indicates that RCS radioactivity concentration is greater than 300 ItCi/gm doseequivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodinespikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this conditionindicates that a significant amount of fuel clad damage has occurred., it represents a loss of the FuelClad 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 ~hathit is recognized that sample collection and analysis of reactor coolant with highly' elevated activit;yevels c.uld reqUire sever.al hour-s to complete. Non.etheless, a sample rlted threshold is includedas 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 ooelantnoble gas and iodine inivenitory., with RCS radioactivity concentration equal to 300 ~t~i/igin dozeequiv~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~gfidose 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 ar-adiationi moniitor- Feading componient.,Add this: p~aragraph (or- similarwding to the Basis if the threshold includes a sample anialysiscomponfenltý.1t is recognfized that sam~ple collectioni and anialysis of reactor- coolant with highly elevatedno 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-1- a a117P U M V IM" iv 11111 lPage MR-FUEL CLAD BARRIER THRESHOLDSNot Applicable (included for numbering consistency)4-40-5. Other IndicationsLoss and/or Potential Loss 5.AThis subcategory addriesses Other Site specitic thresholdS that maly be icl-:ded to indicate lass or potentialless of the Fu.. Clad barrier based en plant specific d en char-a.teriSticS nOt con.Sider.ed in !he gef-leig-'idanee.N/ADeVeloper Notes:Developers should determine& if other- reliable inidieator-s exEist to ev~aluate the status of this fission proeductbarrier (e.g., r-eview accidenit analyses descrFibed in the site Final Safety Analysis Repor-t, as, upidated). Thegolis to identifY aHny unqeorstpecific indicationis that will proemote timely~ and accurate assessmentofbrirStatus.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 Judgmenti 1,di the othertiisfitti Mak f~ u togE-et~ile rcii'eLoss 6.AThis threshold addresses any other factors that may be used by the Emergency Director in determiningwhether the Fuel Clad Barrier is lost.Potential Loss 6.AThis threshold addresses any other factors that may be used by the Emergency Director-in determiningwhether the Fuel Clad Barrier is potentially lost. The Emergency Director should also consider whether ornot to declare the barrier potentially lost in the event that barrier status cannot be monitored.112 1P a iz e IFWR-FUEL CLAD BARRIER THRESHOLDSDceveoper- Notes:No~e113 P P ... e PWR-RCS BARRIER THRESHOLDSThe RCS Barrier includes the RCS primary side and its connections up to and including the pressurizersafety and relief valves, and other connections up to and including the primary isolation valves.RCS or SG Tube LeakageLoss l.AThis threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic ormanual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents aloss 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 beinto any location -inside containment, to the secondary-side (i.e., steam generator tube leakage) oroutside of containment.A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injectionis considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside ofcontai-nent, the declaration escalates to a Site Arca Em..geneySITE AREA EMERGENCY-_since theContainment Barrier Loss threshold I.A will also be met.Potential Loss l.AThis threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizerlevel 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 crewsupervision, directs that a standby charging (makeup) pump be placed in service to restore and maintainpressurizer 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 beinto any location -inside containment, to the secondary-side (i.e., steam generator tube leakage) oroutside of containment.If a leaking steam generator is also FAULTED outside of containment., the declaration escalates to a s4eAfea EmeFg-eneySITE AREA EMERGENCY- since the Containment Barrier Loss threshold 1.A will alsobe met.Potential Loss 1..BIntegritv -Red entry conditions This condition indicates an extreme challenge to the integrity of the RCSpressure boundary due to pressurized thermal shock -a transient that causes rapid RCS cooldown whilethe 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.A1141 Pa, e 4WR-RCS BARRIER THRESHOLDSDepe~nding upon char.ging ptl~m)tew.oum ... ...-pd+~44~eteF4el.epers-i4+ay-Usean RCS leak!rate valae of 50 gpmo a pporit site specifie v'alue. a-, anm altei-nate Potential Lossthr-eshold. Vf used, the thr-eshold wording shouild refleet 411a4 the dtrintc of the leak rate valueexcludes nor-mal redctien in RCSiventory (e.g.. b'. !he letdown systemn or RCP sea! lealzoff)-.Potential Loss 1.14Eniter the site specific indications that define n;, eXtr-eme challenigeto the integrit4y of the RCS piessurboundary dule to pressurized thermial shook a itransient that cauises r-apid RCS eooldown while the RCSis, in N4ode 3 OF higxher (i.e., hot anld pr-essur-ized). These ,vill typic-ally be parameter-s ansd valuestaWOUld 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 anexcessive RC--od-own may not be enitered or- im:medi atelyv ex.,ited if RCS pressbtwe is bel ow a certainFor- 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 theSpecific 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/enqtrydecisioni points or- condition validation crziter-ia. For. exiample. a thrieshold might r-ead "RCs lIntegrfity (P)Red etycniin e ihRSpesr>30pi.2. Inadequate Heat RemovalLoss 2.AThere is no Loss threshold associated with Inadequate Heat Removal.Potential Loss 2.AHeat Sink -Red entry conditions met (- NR level in allA," SG < 14% 134%] OR pressure in at leastone 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 steamgenerators (i.e., loss of an effective secondary-side heat sink). This condition represents a potentialloss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions duringwhich operators intentionally reduce the heat removal capability of the steam generators; duringthese conditions, classification using threshold is not warranted.Meeting this threshold results in a SiteArea SITE AREA EMERGENCY because thisthreshold is identical to Fuel Clad Barrier Potential Loss threshold 2.B; both will be met. This conditionwarrants a Site Area Emergency SITE AREA EMERGENCY declaration because inadequate RCS heatremoval may result in fuel heat-up sufficient to damage the cladding and ifnereaseraise RCS pressure tothe point where mass will be lost from the system.Devetop-rNaote-s+lPotewitm l Los11151 Pa ge P3WR-RCS BARRIER THRESHOLDSEnt- 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 wouEdreuie 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 ntsHL,'k',e. ioee.is sheiiL;d e.iisi iLt meiLiiiii nE .tfifIS 10 8 t ie Sallie HS, er StL, f.. r i ImiJlafto. Heat 51flK medei11,'fconditions met" in accordanee with the guH.3.. RCS Activity / Containment RadiationalG- 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 tOan instantaneous r-elease of all r-eactor cooelant miass into the conitainmnt asuig that reactor coclanitactivity equals Tech.nical Spe.ification all..abvih.le lmits. The calculation resu lt..as that cspetainment fi-adiation motnitor-s wouild bredn'10mR'hr. These monitors have an overage bag-kground feading op1- Wiffh dule to thel _9ene f a kep -lv"sore he vaRlue of- R1hr ;;As; peLectead bcuetha4tRCS. F80r ap dirox fratemy 40 minutoe fRlloyine a lp p.. roim l 40 mne , RT 8050 and RTreadingsn texpe ete to .............. a apct -n ue ufei n h h-sel a i h udb iif 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 specifedfor Fuel Clad Barrgier Loss threshold 3.A s~ince it indicates a loss of the RCS BarrFier only'. Not ApplicablePotential [..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.lantnoble gas and iodine inven.tory,, with R .S ,.tivit .at Tech.ical Specificalion, allowable limits, in'to thlecontainment atnfs4t-ere. U"sin:g RC-Saetkity atTechnical 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 containmentradiation 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 iradkition1161 Page PIWRRCS BARRIER THRESHOLDSDeveloper Notes, 44 LtePgtenyial Bpss4. Containment Integrity or Bypasss 5 An dppvmnpH+RiwA*e4 Ac4(w-eee4FNot Applicable (included for numbering consistency)5. Other IndicationsLoss and/or Potential Loss 5.AT-his Subeategoryý addresses other site specific thr~esholds that may, be ineluided to iidicatc l05s orF potenltialless of the RCS barriier based en plant speeifie design eharacter-isties not c-nsidefr in the gen.ericguidance. N/ADcvelopcr 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 producetharri.e" (e.g.. re\','".. acden a....es des.ibed in; l the, ,ite inal, ysi Fer. as updated). 'h41egEoal iS to idefitif,' any' un ~e0.stpeeific- inidieations that will pRo~note timely anid accurate as'sessmnentof barrier stats.Any' added thresholds should r-epresenit appro.ximately the same r-elative thveat io the-4+arries- as the otherthresholds in this column. Basis infor-ma4ion fori the other thr-esholds, may be used to gauge. the r-elaiivebarrier threat leve-.Variables used to monitor for the significant breach or the potential significantbreach 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 operatorto detect conditions indicative of significant failure of any of the three fission product barriers or thepotential for significant failure of these barTiers. Although variables selected to fulfill Type Cfunctions may rapidly approach the values that indicate an actual significant failure, it is the finalsteady-state value reached that is important. Therefore, a high degree of accuracy and a rapidresponse time are not necessary for Type C information display channels. Type C variables are foundin UFSAR Table 7B.6-1.6. Emergency Director JudgmentLoss 6.AThis threshold addresses any other factors that may be used by the Emergency Director in determiningwhether the RCS Barrier is lost.Potential Loss 6.AThis threshold addresses any other factors that may be used by the Emergency Director in determiningwhether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or notto declare the barrier potentially lost in the event that barrier status cannot be monitored.1171 Page I PW1-R-CONTAINMENT BARRIER THRESHOLDSD~oper- Notes!NetieThe Containment Barrier includes the containment building and connections up to and including theoutermost containment isolation valves. This barrier also includes the main steam, feedwater, andblowdown line extensions outside the containment building up to and including the outermost secondaryside 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 EmergencyGENERAL EMERGENCY.I1. RCS or SG Tube LeakageLoss ILAThis threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outsideof containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordancewith the thresholds for RCS Barrier Potential Loss I.A and Loss 1.A, respectively. This conditionrepresents a bypass of the containment barrier.FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarilydependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steamgenerator is decreasing uncontrollably [part of the FAULTED definition] and the fati--ed-FAU.LTEDsteam generator isolation procedure is not entered because EOP user rules are dictating implementation ofanother procedure to address a higher priority condition, the steam generator is still consideredFAULTED for emergency classification purposes.The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that mayrequire an emergency classification. Steam releases of this size are readily observable with normalControl Room indications. The lower bound for this aspect of the containment barrier is analogous to thelower bound criteria specified in IC SU3 for the fuel clad barrier (i.e., RCS activity values) and IC SU4for the RCS barrier (i.e., RCS leak rate values).This threshold also applies to prolonged steam releases necessitated by operational considerations such asthe forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown theplant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in asignificant and sustained release of radioactive steam to the enviromnent (and are thus similar to aFAULTED condition). The inability to isolate the steam flow without an adverse effect on plantcooldown meets the intent of a loss of containment.Steam releases associated with the expected operation of a SG power operated relief valve or safety reliefvalve do not meet the intent of this threshold. Such releases may occur intermittently for a short period oftime following a reactor trip as operators process through emergency operating procedures to bring theplant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with theunexpected 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-sidesystem component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases donot constitute a loss or potential loss of containment but should be evaluated using the RecognitionCategory ArR ICs.1181 Page PWR-CONTAINMENT BARRIER THRESHOLDSThe emergency classification levels-EMERGENCY CLASSIFICATION LEVELS resulting fromprimary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarizedbelow.Affected SG is FAULTED Outside of Containment?P-to-S Leak RateYesNoLess than or equal to 25gpm (a,,l-hal+*-,erC1 Developer Notes)Greater than 25 gpm (-oother value per SUt4De;'clOpfrf otesRequires operation of astandby charging pump (RCS BarrierPotential Loss)Requires an automatic ormanual ECCS (SI)actuation (RCS BarrierLoss)No classificationU+usual Event UNUSUALEVENT per SU4Site Area Emergency SITEAREA EMERGENCY perFS1Site Area Emer'gen:cy SITEAREA EMERGENCY perFS1No classificationUnisual Event UNUSUALEVENT per SU4Ale4 ALERT per FA IAle4 ALERT per FA IPotential 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 dumpvalve ..r..ter.a.popr... " i. e specific ......Developer-S may include anl additional site 'specilic thre-Sho0ld(s') to at-ddress proelonged steami releasesnecessitated by oper'ational conisider-ations ifANOPS 8r FOPS cou1ld Fectblile that a leakingeiorRUPTPURE-Dsteami genefatef- be utsed to support plant eooldevwn.eveopers tabe intOu tSer. aids o. vallb 4 erloea4ions ;v wiin theirh asi docum t.2. Inadequate Heat RemovalLoss 2There is no Loss threshold associated with Inadequate Heat Removal.119 19P a a e PAIR CONTAINMENT BARRIER THRESHOLDSPotential Loss 2.ACore Cooling -Red entry conditions met for 15 minutes or longer. This condition represents anIMMINENT 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 havebeen a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restoreadequate core cooling is not effective (successful) within 15 minutes, it is assumed that the eventtrajectory 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 decreasingand/or if reaetor vessel vRCS level is increasing. Whether or not the procedure(s) will be effectiveshould be apparent within 15 minutes. The Emergency Director should escalate the emergencyclassification 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 canarrest core degradation in a significant fraction of core damage scenarios, and that the likelihood ofcontainment failure is very small in these events. Given this, it is appropriate to provide 15 minutesbeyond 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 cnerneeiingthe 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.xitthermocupe readi thresholds.Potential LaOss 2.A. 1PEnter site specific er-itria. requir-i-g;3ntry kinte a core cooling restor-ation Proedurc OF pOromptimplementation 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, enterthe parameters and values uised ini the Cor-e Coolingf Red Path.wes tin zh use ERG Plant4Dev~!)eloeS shouE!dld osiderinci!luding a thr-eshold the same as. Or Simfi!ar- to, "Care Coo! ling Red enitryconlditions Met for 1 H! minutesi OF longer" ini accor-dace withi the guidaniee at the front of this section.1201 Page PWR-CONTAINMENT BARRIER THRESHOLDS3. RCS Activity / Containment RadiationLoss 3There is no Loss threshold associated with RCS Activity / Containment Radiation.Potential Loss 3.A. IThe radiation readings for the containment high range area monitors (RT-8050 and RT-8051) correspondmonitor reading eorresponds to an instantaneous release of aEll eaeeor coolat Rs-the radioactivematerial 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 onCalculation No. STPNOC013-004 Rev.2. The threshold values used were conservatively rounded within2% of the calculated values to ensure the values were readily assessable. This level of assumed fuel cladfailure is well abe-,e-bevond that used to determine the analogous Fuel Clad Barrier Loss and RCS BarrierLoss thresholds. Temperature induced current (TIC) limitations are not applicable to the ContainmentBarrier Potential Loss threshold 3.A.1 because the expected radiation dose for this event overwhelms theTIC 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 majorrelease of radioactivity requiring offsite protective actions. For this condition to exist, there must alreadyhave been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat thiscondition as a potential loss of containment which would then escalate the emergency classification levelEMERGENCY CLASSIFICATION LEVEL to a General Emergency GENERAL EMERGENCY.Potential Loss 3.A.2The HATCH MONITOR is located outside containment and is the back-up monitor to the containmenthigh range monitors (RT-8050 and RT-805 1). The HATCH MONITOR threshold value is based onCalculation No. 03-ZE-003. This value corresponds to the calculated containment high rangc monitorreadinas for Containment Barrier Threshold Potential Loss 3.A. 1.Deveoper- Notes:Potential LOss 3.AN UREG 12_. So,'cc L:otic:,; Dr"'ing- Responsef ,qiq,5e eXe I -ower P!kc.ni Ieeiedk;i-sp'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 and4mspe: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 BypassLoss 4.A121 IPa,-e MIWRCONTAINMENT BARRIER THRESHOLDSThese thresholds address a situation where containment isolation is required and one of two conditionsexists as discussed below. Users are reminded that there may be accident and release conditions thatsimultaneously 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 likelyexceeds that associated with allowable leakage (or sometimes referred to as design leakage). Followingthe release of RCS mass into containment, containment pressure will fluctuate based on a variety offactors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable dropin containment pressure. Recognizing the inherent difficulties in determining a containment leak rateduring accident conditions, it is expected that the Emergency Director will assess this threshold usingjudgment, and with due consideration given to current plant conditions, and available operational andradiological 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 leakagefrom a penetration and the other is leakage from an in-service system valve. Depending upon radiationmonitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted inthe figure.Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence oftwo FAULTED locations on a steam generator where one fault is located inside containment (e.g., on asteam or feedwater line) and the other outside of containment. In this case, the associated steam lineprovides 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 beminor radiological releases associated with allowable (design) containment leakage through variouspenetrations or system components. These releases do not constitute a loss or potential loss ofcontainment 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 radioactivematerial 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 withthe 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 dropin containment pressure.Refer to the top piping run of Figure 9-F-344 in Addendum 3. Containment Integzrity or BypassExamples. In this simplified example, the inboard and outboard isolation valves remained open after acontainment isolation was required (i.e., containment isolation was not successful). There is now anUNISOLABLE pathway from the containment to the environment.The existence of a filter is not considered in the threshold assessment. Filters do not remove fissionproduct noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loadingbeyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/highhumidity in the release stream.Leakage between two interfacing liquid systems, by itself, does not meet this threshold.I1221 P a e PWR-CONTAINMENT BARRIER THRESHOLDSRefer to the bottom piping run of Figure 9-F-43. In this simplified example, leakage in an RCP sealcooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would bedetected by the Process Monitor. If there is no leakage friom the clesed water c.eling Component CoolingWater system to the Auxiliary Building, then no threshold has been met. If the pump or system pipingdeveloped a leak that allowed steam/water to enter the Auxiliary Building, then threshold 4.B would bemet. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by anyof 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 beminor radiological releases associated with allowable (design) containment leakage through variouspenetrations 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 constitutea 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 assessedusing Loss Threshold l.A.Loss 4.BContainment sump, temperature, pressure and/or radiation levels will inef-easerise if reactor coolant massis leaking into the containment. If these parameters have not i'ereasedrisen, then the reactor coolant massmay 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 thatthe RCS mass is being lost outside of containment.Unexpected elevated readings and alarms on radiation monitors with detectors outside containmentshould be con'oborated with other available indications to confirm that the source is a loss of RCS massoutside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside ofcontainment may not -neFeaserise significantly; however, other unexpected changes in sump levels, areatemperatures or pressures, flow rates, etc. should be sufficient to determine if RCS mass is being lostoutside 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 coolantin the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage couldbe 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 containmentmust be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold I.A to bemet.Potential Loss 4.AContainment -Red entry conditions met (containment pressure > 56.5 tSIG). If containment pressureexceeds 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 andFuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a Site Aiea123 1P a g e PWR CONTAINMENT BARRIER THRESHOLDSEmergeney -SITE AREA EMERGENCY and General Emergency GENERAL EMERGENCY since thereis now a potential to lose the third barrier.Potential Loss 4.BThe existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogenconcentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limitI4!4.}). Ahydrogen burn will raise containment pressure and could result in collateral equipment damage leading toa loss of contaim-nent integrity. It therefore represents a potential loss of the Containment Barrier.Potential Loss 4.CThis threshold describes a condition where containment pressure is greater than the setpoint (9jPSIG.P-P-4) at which containment encrgy,.heat) removal systems are Containment Spray is designed toautomatically 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 haveautomatically 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 adegraded 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.edMO94ito1r alrmS OF readings may also be included.Potential Less !.,kThe 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. tilepressure value in Potential Inoss 4.A is that used for- the Containment Red Path. if the ConitainmentCPSS4 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 4.13D)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-thashould b- .pe' designi the containment press-ure setpoi,," is r. .,I.,,.. 1. desired. specificconditioni indEicatios5 51u10h as paramete- N'alues cani also be enter-ed (e.g., a containment4 spray h1ow r-ateless 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 Indicationscflmm;;I M, mi~ LaCreIaoi 2Ce 3Aitil tile '2L1UidaiCC atL tile K1r11[ Bi- [[IS :;eet!1.Loss and/or Potential Loss 5.A124 1Pa e PW1 CONTAINMENT BARRIER THRESHOLDSThis su,1bcat-g.oy addr-esses other site specitic thesh'-lds that ma;,' be-included to indicate lo,'ss ort peM+t-Ua10S9 Of the COntainmen~lt barrier basied an plant specific desig n chairacteristics niat coansider~ed iii the genler-iguidance. N!ADevlper- NoteI oss and/or Potential L.,oss 5.A6. seemergency iretnrpovide f venting othe containment as a means of prevtentingaTaisthreshid ailureaLsse a threrhl fcosthatl ay included for the otimerncy bDriretr. Tin dtrerhoininglbe met as soeo as such venting is loINsENT. Containment ventin as part of rec-er s i7classified 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 productbarrier. (e.g., review. accident analyses descr-ibed in thiesite.Final Sft'AnlisRepor t. as updated). Thegoal is te identifal any unie or site specific indications that will pr6mote timel and accurate assessmentof barrier status.AHN' added thr-eshodS Should! repre.sent approximately the samie r-elative thrfeat to the harriier, as the othierthrshld inths olun.BaiS in41-foration for1 the Other threshOlds may' be us-ed to gauge the@ relativebar-rier threa.: level.6. Emergency Director JudgmentLoss 6.AThis threshold addresses any other factors that may be used by the Emergency Director in determiningwhether the Containment Barrier is lost.Potential Loss 6.AThis threshold addresses any other factors that may be used by the Emergency Director in determiningwhether the Containment Barrier is potentially lost. The Emergency Director should also considerwhether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.Developer- Notes:Noee125 1Pa e PAWR-CONTAINMENT BARRIER THRESHOLDS:126 1Page Figure 9-F-43: VWR-Containment Integrity or Bypass ExamplesRCP SealCoolingNOTES: Only Supplemental Purge is a filtered release and STPEGS Component Cooling Water is equivalent to Closed Cooling Water127 1Page 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANTSAFETY ICS/EALSTable H-i: Recognition Category "H" Initiating Condition MatrixUNUSUAL EVENTHU1 ConfirmedSECURITY CONDITIONor threat.Op. Modes: A4lALLALERTSITE AREAEMERGENCYGENERALEMERGENCYHAI HOSTILE ACTIONwithin the OWNERCONTROLLED AREA orairborne attack threatwithin 30 minutes.Op. Afodes: A4/ALLHS1 HOSTILE ACTIONwithin the PROTECTEDAREA.Op. Modes: A//ALLHG1 HOSTILE ACTIONresulting in loss of physicalcontrol of the facility.Op. Modes: A4/ALLHU2 Seismic event greaterthan OBE levels.Op. Modes: A//ALLHU3 Hazardous event.Op. Modes: A44ALLHU4 FIRE potentiallydegrading the level ofsafety of the plant.Op. Modes: Note:(Q SeeSA9orCA6 for escalationof these eventsHA5 Gaseous releaseimpeding access toequipment necessary fornormal plant operations,cooldown or shutdown.Op. Mllodes: A-1ALLHA6 Control Roomevacuation resulting intransfer of plant control toalternate locations. Op.Modes: ,4/4ALLHA7 Other conditionsexist which in thejudgment of theEmergency Directorwarrant declaration of anAleAALERT.Op. Alodes: A-tALLHU7 Other conditions existwhich in the j udgment ofthe Emergency Directorwarrant declaration of an(N.O)UE UNUSUALEVENT.Op. AModes: A4/ALLHS6 Inability to control akey safety function fromoutside the Control Room.Op. Modes: A4-ALLHS7 Other conditions existwhich in the judgment ofthe Emergency Directorwarrant declaration of aSite Area Emergencyg4TF=SITE AREAEMERGENCY.Op. Modes: A//ALLHG7 Other conditionsexist which in thejudgment of theEmergency Directorwarrant declaration of aGeaneral Em~ergencyG ERNAE GENERALEMERGENCY.Op. Modes: A/.ALL1281 Page HU1ECL: of U....s..al UNUSUAL EVENTInitiating Condition: Confirmed SECURITY CONDITION or threat.Operating Mode Applicability: AI-ALLS Emergency Action Levels: (1 or 2 or 3)( L) A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by -spe4fiesecurity 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 thusrepresent a potential degradation in the level of plant safety. See*4itv e'e ts-SECURITY EVENTS whichdo not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10CFR § 50.72.. Security events SECURITY EVENTS assessed as HOSTILE ACTIONS are classifiableunder ICs HAl., HSI and HGI.Timely and accurate communications between Security SkiP4-Force Supervision and the Control Room isessential for proper classification of a security-related event. Classification of these events will initiateappropriate threat-related notifications to plant personnel and OROs.Security plans and terminology are.based on the guidance provided by NEI 03-12, Templatef/r theSecuriv., 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 arethe 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 thenature of Safeguards and 10 CFR § 2.39039 information.EAL #2- addresses the receipt of a credible security threatCREDIBLE SECURITY THREAT. Thecredibility 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 HeadquartersOperations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The statusand size of the plane may also be provided by NORAD through the NRC. Validation of the threat isperformed 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 notincorporate Security-sensitive information. This includes information that may be advantageous to apotential 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 bevia 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 becausethe Acting Security Manager position is not normally activated until after an UNUSUAL EVENT hasbeen declared, and the Security .Manager position is not normally activated until after an ALERT has beendeclared. However, reporting by the Acting Security Manager or Security Manager was included in theevent these positions are staffed under unusual circumstances.REFERENCEs:1. OERPOI-ZV-SH03, Rev. 12. Acting Security ManagerOER.P0 I -ZV-TS08. Rev. 16. Security Manager3. 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 forThe (site Speeific prOCedur~e) is the procedur-e(s) uised by Controel Roefm and/lor Security perSENnfel todeter-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 not4eorort securit~; sens~itive in1formation. This iniekudes inforimation thiat may be advanitageous to aPotential adversary", suhas the partIiculars, concerning a specific thr-eat at- threat locationi. Securiitysensitive 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-mberede-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.A130 1Pa e HU2ECL: Nat-fieatitn of U..us.al E9.N UNUSUAL EVENTInitiating Condition: Seismic event greater than OBE levels.Operating Mode Applicability: AIIALLEamnpe-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.eOE-4mn'ASE1SMIC EVENT" alarm in U nitI Control Room (L-ampbox 9M01, Window E-8)OR2. Control Room personnel feel an actual or potential seismic event.ANDb. The occurrence of a seismic event is confirmed in manner deemed appropriate by the ShiftManager or Emer.,ency Director.Basis:This IC addresses a seismic event that results in accelerations at the plant site greater than those specifiedfor an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a SafeShutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures andcomponents; however, some time may be required for the plant staff to ascertain the actual post-eventcondition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessaryto perform walk-downs and inspections, and fully understand any impacts, this event represents apotential 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 ,), itdoes 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., SeismicEvent 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 ControlRoom personnel and which can be readily assessed. Indications available outside the Control Roomand/or which require lengthy times to assess (e.g., processing of scratch plates or recorded data) shouldnot be used. The goal is to specify indications that can be assessed within 15-minutes of the actual orsuspected seismic event.The EAL 1 .b- statement is included to ensure that a declaration does not result firom felt vibrations causedby a non-seismic source (e.g., a dropped heavy load). Event verificatian with external sou-rces souald notbe fneeessary d uing or- follwinlg ani OBE. E~arthquakes of this miagnituide shouild be r-eadily felt by. oni sitepers.nnel and recognized as a seismi. event (eg.typial lateral accelerations .inw ex.ess 4f0.OSg). The1311 Pa e Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call tothe USGS, check internet news sources, etc.); however, the verification action mnust not preclude a timelyemergency declaration. It is recognized that this alternate EAL wording may cause a site to declare anUNUSUAL EVENT while another site, similarly affected but With readily assessable OBE indications inthe Control Room. may not.Depending upon the plant mode at the time of the event, escalation of the emergency classifieation levelEMERGENCY 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 hasexperienced an OBE. The Seismic Event Alarm setpoint is 0.02g in the vertical or horizontal position andthe station design basis value for an OBE is 0.05oý. Since the Seismic Event alarm is set akt less than halfof the OBE value, it cannot be used as the sole threshold value for determinin. whether or not STP hasexperienced an OBE.STP has implemented the alternative EAL described in NEI 99-01 Developer Notes in conjunction withusing the installed indication. EAI,-. b. allows the Shift Manacer or Emergency Director to determine ifa seismic event has taken place. taking into consideration the Seismic Event alarm, Control Roompersonnel feeling an actual or potential seismic event and other indications deemed appropriate.REFERENCES:I. OPOP04-SY-000 I. Rev. 8, Seismic Event2. 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 thein.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 ConltrolRo00M per-sonnel and which can be ix'adily assessed. indications available outside the Control Roomand/orWlich require lengthy times--to-ssss , r d a -suldREAt be use.d. The goal is to specify indications that Can be assessed within 15 minutes of the actual orspcedseismic event.For site-s that do not have readily assessable OBE inldicatiOnS within the Control0 RoomI, de-Velopers shou~lduse 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 ShiftManager 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 maseek 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 :ecognizedthat this alter.nate [AL.= w.ording may cause a site to.declare an u ............e.t .wrhile ano.er. site, similarlyaftereted but A ith Feadik assessable "PIE indiecttiwis iii the C.entfol Room fflay not.I132 1 P a g e Nwheli a secismie FaanitaringqsytemcmECL Assil-nrent Attributes: -3. 1. lAapahle of deteting an OBk is wt of geF,,iee fer inainteiiaiie Of133 IP a e HU3ECL: Natiticati.n of Unus..al Event UNUSUAL EVENTInitiating Condition: Hazardous event.Operating Mode Applicability: A-I4ALLExanpie 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 vehiclebreakdowns 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 electricalisolation 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 eventinvolving 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 fromaccessing the site via personal vehicles.(3) (Site spe.if. , list of .atu.al or " hazard Predicted or actual breach of MainCooling Reservoir retaining dike along North Wall(4sL)_Basis:This IC addresses hazardous events that are considered to represent a potential degradation of the level ofsafety 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 aSAFETY SYSTEM component due to water level or other wetting concerns. Classification is alsorequired if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEMcomponent from its power source (e.g., a breaker or relay trip). To warrant classification, operability ofthe 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 sufficientmagnitude 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 andsignificant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples ofsuch an event include site flooding caused by a hurricane, heavy rains, up-river water releases, damfailure, etc.. or an on-site train derailment blocking the access road. This EAL is not intended apply to134 1 P a 2 e routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to moresignificant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding aroundthe Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in2011.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 acredible hazard and analyzed in the STPEGS UFSAR..Escalation of the emergency classification level EMERGENCY CLASSIFICATION LEVEL would bebased on ICs in Recognition Categories AR, F, S or C.iHU3: EAL-i, EAL-2, EAL-3, EAL-4 Selection Basis:N/AREFERENCE:I. STPEGS ULI'SAR. Section 3.4.1, Flood ProtectionDeveloper Notes:.he "Site specitic list of natu'al or technological hazard events" sheuld E!in.de other events that mia- be aprecursor to a moe.. significant eve..t or cndition, and that are appri.. priate to the site .ation aNotwithstanding fl"..thle ce,,ts spe.ifie.. i..cluded as EAks above, a "Site sp.cifie list 'f natural ortechnologic-al hazar-d evenits" need not inckludesor lived events for wh.4ich the exNtent of the damage andthe .es.lti.n c-ne ...... es can. be deter'minied w.ilhin a r "elatively short ti.. e frame. in these cases. adamage 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 andimpleffmentation 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 anldstructL 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. hiaddiiti-en, the--eeeurrenee or- effects of the e~venit mw,- be CPECL Assignment4,÷ "1 1 1 A ... I 2 1 1 1-')6X3mr- :ý- 0 Ing1351 Page HU4ECL: -f ... us4T al E.. eit .UNUSUAL EVENTInitiating Condition: FIRE potentially degrading the level of safety of the plant.Operating Mode Applicability: A-IALLExta-mple Emergency Action Levels: (1 or 2 or 3 or 4)Note: The Emergency Director should declare the Unuis-ial Eveit-UNUSUAL EVENT promptly upondetermining 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 detectionindications:* Report firom the field (i.e., visual observation)* Receipt of multiple (more than 1) fire alarms or indications* Field verification of a single fire alarmANDb. 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-NTable HA4: Plant Roorns/Areaso 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'4Turbine Genierator BuikldingFuel Handling BuildingP. "ne.t:- C":'ii.n h::n.rn R:i!,1linoR-ildilw.........E&Sential Cool ing Water Intake Str'ctureIsolation Valve CuibicleDiesel Genlerator BuildingCirculIAtigater- Inake Str-ucture(2) a. Receipt of a single fire alarm (i.e., no other indications of a FIRE).ANDb. 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)witehlyardTur-bine Gjener-ator BuildingMeehanieal,lEleetrieal ~A~teiliai-y BuildingFuel ,andlin ,uIl--ingReaetor: Containment BuildiriigEssential Cooling Water- Intake StructureIsolationi Valve CubieleDiesel Generator BuildingCirculatinig Water- hitake Struttur~eANDc. 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 orindication.(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 toextinguish.Basis:This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation ofthe level of safety of the plant.EAL #1The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that arereadily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of aFIRE 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 initialalarm, indication, or report was received, and not the time that a subsequent verification action wasperformed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm,indication or report.EAL #2This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., provedor disproved) within 30-minutes of the alarn. Upon receipt, operators will take prompt actions to confirmthe validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the timethat 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 aspurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify thevalidity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIREexists; however, after that time, and absent information to the contrary, it is assumed that an actual FIREis in progress.If an actual FIRE is verified by a report friom the field, then EAL #1 is immediately applicable, and theemergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarmis verified to be due to an equipment failure or a spurious activation, and this verification occurs within30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration iswarranted.EAL #3In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant or ISFSI PROTECTEDAREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. 44i6basis extends to a FARE- ocellrr4in wihntePOETD-L' fa SS oae udeth anPROTECTED rRg. nt with. a POES! ousd r e&,-, ....... .,. ., the plant Pv.teetedrea]EAL #4If 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., alocal town Fire Department), then the level of plant safety is potentially degraded. The dispatch of anoffsite firefighting agency to the site requires an emergency declaration only if it is needed to activelysupport 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 RAppendix R to 10 CFR 50, states in part:Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components importantto safety shall be designed and located to minimize, consistent with other safety requirements, theprobability and effect of fires and EXPLOSIONS."When considering the effects of fire, those systems associated with achieving and maintaining safeshutdown conditions assume major importance to safety because damage to them can lead to coredamage resulting from loss of coolant through boil-off.Because fire may affect safe shutdown systems and because the loss of function of systems used tomitigate the consequences of design basis accidents under post-fire conditions does not per se impactpublic safety, the need to limit fire damage to systems required to achieve and maintain safeshutdown conditions is greater than the need to limit fire damage to those systems required tomitigate the consequences of design basis accidents.In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of l-hour firebarriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train1381 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-hourtime period.Depending upon the plant mode at the time of the event, escalation of the emergency classifieation levelEMERGENCY 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 Program2. STPEGS UFSAR. Rev. 16,_Section 7.4. Systems Required for Safe Shutdown139 1 P a g e The '"ite speei fie lit 4f planit r-oomsS YSTEM equipmeiit.As nioied in the FAIL.- anid Basis see!Prote~'edkie -:1-1-1A A ;. 1 .- , .2 1 ioi. areas" Should spe.it.. those rooms at- areas that contain SAFETYion,. include thle term IsfEsl if the site has an ISFSI oultqide thle PlantttF HtE!S: ...AI140 e HU7ECL: Notification of Unusual Event UNUSUAL EVENTInitiating Condition: Other conditions exist which in the judgment of the Emergency Director warrantdeclaration of a (NOýUE.Operating Mode Applicability: AI4ALLEAitniF!e-Emnergency Action Levels:(1) Other conditions exist which in the judgment of the Emergency Director indicate that events arein progress or have occurred which indicate a potential degradation of the level of safety of theplant or indicate a security threat to FAC1LITYfeaci-4-y protection has been initiated. No releasesof radioactive material requiring offsite response or monitoring are expected unless furtherdegradation of safety systems occurs.Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declarationof 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/AREFERENCES:N/A141 1Pa e HA1ECL: A4eq4 ALERTInitiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborneattack threat within 30 minutes.Operating Mode Applicability: AI1ALLExample-Emergency Action Levels: (I or 2)(1) -....A HOSTIL6E ACTION is eeecurringe or has ecculfed within the OWNER CONTROLLED AREAas repaitcd by the (siteSj,.o, spec- u i-s S v .c S-pep.'iszp OR ActingS t Mane. A HOSTILE ACTION is occurring or has occurred within the OWNERCONTROLLED AREA as reported by ANY of the following personnel in Table HI:Table HI: Securill Supervision" Security Force Supervisor* Acting Security Manager* Security Manager14A 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 AREAor notification of an aircraft attack threat. This event will require rapid response and assistance due to thepossibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and stafffor a potential aircraft impact.Timely and accurate communications between Security Shift Supervision and the Control Room isessential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for theSecurity Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Spe-ý4FV'o c! ... S...... Jn..t.allatioh" INDEPEAVDE.NT SPENT FL/EL STOt-4GE INSTALLA TION SecurityProgram].As time and conditions allow, these events require a heightened state of readiness by the plant staff andimplementation 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 tobe 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 arenot a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small142 1P a R e aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of eventsis 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 OWNERCONTROLLED AREA. This inludes a.'..a.tio. afaifSt all .S.SoI that is l..ated outi. de theplant AREA.EAL #2 addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time iswithin 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in atimely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is metwhen 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 involvesan 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 CONTROLLEDAREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notificationby an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federalagency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one basedon 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 notincorporate Security-sensitive information. This includes information that may be advantageous to apotential 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 -wau4LEVELwould 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 ORActing Security Manager were included because either of these positions could be activated prior tomeeting this EAL. The Security Force Supervisor is a 24-hour position and the normally the ActingSecurity Manager is activated after an UNUSUAL EVENT has been declared. The Security Manager isalso included t4--although this position is normally activated after an ALERT.REFERENCES:I. OERPO I-ZV-Sf 103. Rev. 12. Acting Security Manager2. OERP0 I-ZV-TS08. Rev. 16, SecuritN Manager3. 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 notr tSeouritv sensitive information. This includes infor.mation that malp' be atdantageous to apotetia 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! numbleredreferences 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 schemedefinition for the OWNER CONTROLLED AREA.144 1 P a , e HA5ECL: le# ALERTInitiating Condition: Gaseous release impeding access to equipment necessary for normal plantI operations, cooldown or shutdown.I Operating Mode Applicability: AIIALLI ~ Exam-ple Emergency Action Levels:Note: If the equipment in the listed room or area was already inoperable or out-of-service before theevent 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-ANYof 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 Buildifiý414G134jS0jfttiE)f1 3,701- C-h-1pF-Hel Handline Buildine (FHB)ANDb. Entry into the room or area is prohibited or impeded.I TABL F W/I2: PI-rnt Ar,- P- uridrina ArpzRCB RHR Heat Exchanger Rooms00 rI MAB 51 ft Room 335EAB Roof, MCC 1G8, 4.16KV Switchgear Roomso Ln) EAB 4.16KV Switchgear Rooms145 1 P a g e Basis:This IC addresses an event involving a release of a hazardous gas that precludes or impedes access toequipment necessary to maintain normal plant operation, or required for a normal plant cooldown andshutdown. This condition represents an actual or potential substantial degradation of the level of safety ofthe plant.An A-eA ALERT declaration is warranted if entry into the affected room/area is, or may be, procedurallyrequired during the plant operating mode in effect at the time of the gaseous release. The emergencyclassification 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 EmergencyDirector's judgment that the gas concentration in the affected room/area is sufficient to preclude orsignificantly impede procedurally required access. This judgment may be based on a variety of factorsincluding an existing job hazard analysis, report of ill effects on personnel, advice from a subject matterexpert or operating experience with the same or similar hazards. Access should be considered as impededif 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 theaffected room until Mode 4.* The gas release is a planmed activity that includes compensatory measures which address thetemporary 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 actuallyprevent or impede a required action.An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Mostcommonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces theconcentration 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 firesuppression system in an area. or- t i -ntentional ine.. tiing of .,ntainment (AI3R only).146l P a, e Escalation of the emergency classificationvia Recognition Category AR C or F ICs.level-EMERGENCY CLASSIFICATION LEVEL would beHA5: 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 safeshutdown equipment operation is capable from the Main Control Room (MCR). The plant is able totransition 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 necessaryfollowing entry into hot shutdown (establish Residual Heat Removal shutdown cooling, disable operationof 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 willbe maintained in a cold shutdown condition.REFERENCES:1. OPGP03-ZF-0001, Rev. 26, Fire Protection Progrzam2. STPEGS UFSAR. Rev. 16. Section 7.4. Systems Required for Safe Shutdown3. OPOPO3-ZG-0008. Rev. 56. Power Operations4. OPOP03-ZG-0006, Rev. 54. Plant Shutdown from 100% to Hot Standby5. OPOPO3-ZG-0007. Rev. 71, Plant CooldownDevelopertNot es:The "site specific list Ofplant roomas or areas with entr-y r-elated mode applicability identifled"' shouildope.OC.edures ..sed Ie.r norm"l .plant opei.Aon, co.ldown shud w!n Do not inClude rooms orareas Mn Whlieh ac-tionls 4fa eantingent Or emergency natur'e would be perfor-med (e.g.. an actioni to addre&ssan 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 reOquiredfor eac h room or area.The list should fiot inlude rooms el- areas forw.ih... entry is ..ui.ed solely to perform actions ofanadministr-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 topr-e-kmDe-a-Cntrol Room e-oacuaoie -4uo4e release ofa hazardet.sgas. ma;, inclunde buare not limiited to. capability to draw air- froem multiple air- intakes at deifbent and separ~ate location-ineand outer- atmoespheric boundar-ies, or hiapblt to acquir-e and mainiain positive pr-essur-e w~ithin theIf 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 thataready allowed by Tecfhni.al Specications at te time of the even.t..F:CL AsI enrnn; Attribbites:3.I2B147 1 P a g e HA6ECL: ANe4 ALERTInitiating Condition: Control Room evacuation resulting in transfer of plant control to alternatelocations.Operating Mode Applicability: A44ALLExample Emergency Action Levels:(1) An event has resulted in plant control being transferred from the Control Room to (scite speeiPeie -Auxiliarv Shutdown Panel Basis:This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternatelocations outside the Control Room. The loss of the ability to control the plant from the Control Room isconsidered 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 shutdownlocations. The necessity to control a plant shutdown from outside the Control Room, in addition toresponding to the event that required the evacuation of the Control Room., will present challenges to plantoperators and other on-shift personnel. Activation of the ERO and emergency response facilities willassist in responding to these challenges.Escalation of the erf, egenlcy classification levl E4MERGENCY CLASSIFICATION -wo+d4LEVELwould be via IC HS6.HA6: EAL-1. Selection Basis:The Auxiliary Shutdown Panel (ASP) is identified in OPOP04-ZO-0001. Control Room Evacuation. asthe location where plant control is transferred in the event of a Control Room evacuation.REFERENCES:1. Procedure OPOP04-ZO-0001, Rev. 35. Control Room EvacuationDcvelopcr 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 thecofft+ReýA it .;I, , , -'2 1 ') E)i 11 , 0148 1 P a R e HA7ECL: A-44 ALERTInitiating Condition: Other conditions exist which in the judgment of the Emergency Directorwarrant 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 arein progress or have occurred which involve an actual or potential substantial degradation of thelevel of safety of the plant or a security event SECURITY EVENT that involves probable lifethreatening 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 declarationof an emergency because conditions exist which are believed by the Emergency Director to fall under theemcrgznc.y elazsification l.vcl EMERGENCY CLASSIFICATION deserpenLEVEL description for anAje4-ALERT.HA7: EAL-1 Selection Basis:N/AREFERENCE:N/A149 1 P age HS1ECL: SiteUea Emer.gene.. SITE AREA EMERGENCYInitiating Condition: HOSTILE ACTION within the PROTECTED AREA.Operating Mode Applicability: AI-ALLENample-Emergency Action Levels:(1) A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reportedby the (site spe.ific.se....itoy. shift supe.. ision). -nny ANY of the followine personnel in TableHI:Table HI: Security Supervision* Security Force Supervisoro Acting Security Manager* Security ManalerBasis:This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This eventwill 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 isessential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for theSecurity. Plan, Training and Qualification Plan, Safeguards Contingency Plan [and izde*qeiik- F: .,e! .....ag. 4...a..aiiw--. INDEPENDENT SPENT FUEL STORAGE INSTALL4 TION SecurityProgran].As time and conditions allow, these events require a heightened state of readiness by the plant staff andimplementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site AreaEmergefncy SITESITE AREA EMERGENCY declaration will mobilize ORO resources and have themavailable to develop and implement public protective actions in the unlikely event that the attack issuccessful 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 anotapp4y-4e,-incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILEACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots fromhunters, physical disputes between employees, etc. Reporting of these types of events is adequatelyaddressed 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 notincorporate Security-sensitive information. This includes information that may be advantageous to apotential 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 bevia IC HGI.HSI: EAL-1 Selection Basis:The positions of Security Force Supervisor, Acting Security Manager, and Security Manager wereincluded since any of these positions could be activated prior to meeting this EAL. The Security ForceSquervisor is a 24-hour position, the Acting Security Manager is activated after an Unusual Event hasbeen declared and the Security Manager is activated after an Alert is declared.REFERENCES:1. OERPOI-ZV-SI103, Rev. 12. Acting Security Manager2. OERPOI-ZV-'I"SO8. Rev. 16. Security ManagerDevelepcr- Notes:TPhe (Site spepcifi security "hift super-vision) is the title, of the An 4hif indiv.idual responsible forsupc'.'iionof 4hc ell shift security 4ýforeeEmergelcy' plans and ,impleenetng procedures are public documents; therefore, s should notincrpoateSecuritN,' sensitive infor-mationi. T!his includes, inforemationi that may, be advaiitageeus to ap.t..t.a adversary.. suH. aS ,he P articuflrS conerning a spiic th.e. or:thre" lation. Securitysensitive 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-4retýferenes 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 nELmybworded 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 scemedefinitio-n Por the PROTECTED AREA.ECL Ass-,nment Attrbutes: 3. 1.3.. D151 Pag- e HS6ECL: Site Area Emergency SITESITE AREA EMERGENCYInitiating Condition: Inability to control a key safety function from outside the Control Room.Operating Mode Applicability: A-IALLE*:imple Emergency Action Levels:Note: The Emergency Director should declare the Site Area Emergency SITESITE AREAEMERGENCY promptly upon determining that (site specific num-ber of milutes) 15 minutes has beenexceeded, or will likely be exceeded.(1) a. An event has resulted in plant control being transferred from the Control Room to (site speei-4ereI!!ote shutdon ... el. and l... cntr. l ," atio:s).the Auxiliary Shutdown Panel (ASP).0aefollowinig locations:ANDb. 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 removalTable H2: Safety Functions* Reactivity control* Core cooling* RCS heat removalBasis:This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternatelocations, and the control of a key safety function cannot be reestablished in a timely manner. The failureto gain control of a key safety function following a transfer of plant control to alternate locations is aprecursor 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 ) nAuxiliary Shutdown Panel is based on Emergency Directorjudgment. The Emergency Director isexpected to make a reasonable, informned judgment within (the site specific tim.e ... t..ansr.) 154-0minutes whether or not the operating staff has control of key safety functions friom the remote safeshutdown location(s).Escalation of the emergency classifieation lev'el EMERGENCY CLASSIFICATION LEVEL would bevia 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. asthe location where plant control is transferred in the event of a Control Room evacuation. The 15 minutetimeframe to control the key safety functions is identified as site specific information. The modeapplicability conditioning statement for Table H2 is based on the Technical Specification Operabilityrequirement 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 EvacuationI.2. Technical Specification 3.3.3.5 Remote Shutdown System153 1 P aiz e
- 3. Developer- Notcs:4. The "site specific remote shuttdown paniels and loeal eontroil stationis" arc the panelsancontrtol stationis refei-cnccd in plant proeedur~es uised to cooldown mnd shutdown the planit froema locationi(s) outside the Control Room.5. The "site spec-ific number of minutes" is the tome in which plant control must be (or iecipeeted to be) reestablishied at ani alternate location as descr-ibed in the site specifc &-reresponse 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.B154 1P a g e HS7ECL: Site Area Emergency SITESITE AREA EMERGENCYInitiating Condition: Other conditions exist which in the j udgment of the Emergency Director warrantdeclaration of a Site, Aea fEmegene' SITESITE AREA EMERGENCY.Operating Mode Applicability: A-HALLF*a-vnpk-Emergency Action Levels:_Other conditions exist which in the judgment of the Emergency Director indicate that events arein progress or have occurred which involve actual or likely major failures of plant functionsneeded for protection of the public or HOSTILE ACTION that results in intentional darnage ormalicious 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. AnyANYreleases are not expected to result in exposure levels which exceed EPA eeiP-veAetionG* PROTECTIVE ACTION GUIDELINE exposure levels beyond the .44e nd.ai+y-SITEBOUNDARY.Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declarationof an emergency because conditions exist which are believed by the Emergency Director to fall under theemergen.y classification level EMERGENCY CLASSIFICATION LEVEL description for a Site Area]Fieige+iy SITE AREA EMERGENCY.14S7: gAL-I Selection Basis:N/AREFERENCE:N/A1551 P a e HG1ECL: General Emergeney GENEPLGENERAL EMERGENCYInitiating Condition: HOSTILE ACTION resulting in loss of physical control of the F FACI ITY.Operating Mode Applicability: A-4ALLE1nniple-Ernergency Action Levels:(1) a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reportedby the (site spccifi s.... supe:viwisn). aIn ANY of the following in Table I11:Table HI: Security Supervision* Security Force Supervisor* Acting SecuritV Managero Security ManagerANDb. EITHER of the following has occurred:I_. ANY of the following safety functions in Table H12 cannot be controlled or maintained inMODES 1, 2 or 3 ONLY.Reaetivity efntro~lCorFe eeelin~[Pgl, RP% water- lev~el [XVR]RGS heat r-emovalTable 1-12: Safetv Functions-Reactivity control* Core coolingo RCS heat removalOR-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 4weityFACILITY to the extent that the plant staff can no longer operate equipment necessary to maintain keysafety functions. It also addresses a HOSTILE ACTION leading to a loss of physical control that resultsin 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 waterlevel cannot be maintained.Timely and accurate communications between Security Shift Supervision and the Control Room isessential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEt 03-12, Template for theSecurio, 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 notincorporate Security-sensitive information. This includes information that may be advantageous to apotential 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 alsoincluded since any of these positions could be activated prior to meeting this EAL. The modeapplicability conditioning statement for Table H2 is based on the Technical Specification Operabilityrequirement 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 Manaer2. 0ERP0I-ZV-TS08. Rev. 16, Security Managzer3. Technical Specification 3.3.3.5 Remote Shutdown SystemDevekopei- 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 nMnOrFporate Security sensitive informIfationl. ThiS iclu~kdeS infor-mation thatI may be advantageous to apotential 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 numiberereAef-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 bewor-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-Y158 1 P a g e HG7ECL: General Em'crgency GENERAL EMERGENCYInitiating Condition: Other conditions exist which in the judgment of the Emergency Director warrantdeclaration of a General EmfergnyG ENERAL EMERGENCY.Operating Mode Applicability: A41ALLEmergency Action Levels:(1) Other conditions exist which in the judgment of the Emergency Director indicate that events arein progress or have occurred which involve actual or IMMINENT substantial core degradation ormelting with potential for loss of containment integrity or HOSTILE ACTION that results in anactual loss of physical control of the fFaeilitNTACILITY. Releases can be reasonably expected toexceed EPA Protective Acttion C'ideline PROTECTIVE ACTION GUIDELINE exposure levelsoffsite for more than the immediate site area.Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declarationof an emergency because conditions exist which are believed by the Emergency Director to fall under theemergency' classification level EMERGENCY CLASSIFICATION LEVEL description for a GenefaIE.e.ge.ey GENERAL EMERGENCY.HG7: EAL-I Selection Basis:N/AREFERENCE:N/A1591 P a g e 11 SYSTEM MALFUNCTION ICS/EALSTable S-I: Recognition Category "S" Initiating Condition MatrixUNUSUAL EVENTALERTSITE AREAEMERGENCYSU1 Loss of aI4ALLoffsite AC powercapability to emergencybuses for 15 minutes orlonger.Op. Modes:.OSt,;atdiy ;T, 1,2.3,4SU2 UNPLANNED lossof Control Roomindications for 15 minutesor longer.Op. Modes: Pii&ei-.1,2.3,4SU3 Reactor coolantactivity greater thanTechnical Specificationallowable limits.Op. Modes: Po.ei,fluc'ti;:, 1.2,3,'4SU4 RCS leakage for 15minutes or longer.Op. Modes: P#owe-('.S ..an lb , fiT.- 'l, l .......owL ,2.3.4SU5 Automatic or manual(trip -[4fails to shutdownthe reactor.Op. Modes: P.'wei6ýeoe ...... 2 .. 4SA1 Loss of aI4ALL but oneAC power source to emergencybuses for 15 minutes or longer.Op. Modes: Po;w'i" SJ,.,ar:, up .-, N , IL.,:SA2 UNPLANNED loss ofControl Room indications for15 minutes or longer with asignificant transient inprogress.Op. Alodes: P,.e", Ope;'wi&HTSttup.... -ot Stn..y., H4 ,Shwý&tki1,2,,j4SS1 Loss of a4ALLoffsite and aM4ALLonsite AC power toemergency buses for 15minutes or longer.Op. Modes: P.we"QO .....ioo., .......lq, 4Wt1.2,3,4GENERAL EMERGENCYSG1 Prolonged loss of a4ALLoffsite and alALL onsite ACpower to emergency buses.Op. Modes: Powe+'- St;..t.p,. Hl , ,S ..b, H. :Shmtk"wt-L2_3 4SA5 Automatic or manual (tripPWR]" / r rDal [BW.R") fails toshutdown the reactor, andsubsequent manual actionstaken at the reactor controlee-sa1L-s-anels are notsuccessful in shutting down thereactor.Op. Modes: Powero- en1,2SS5 Inability toshutdown the reactorcausing a challenge to(core cooling [#R]RPV Nrater level [+BR])or RCS heat removal.Op. Modes.: P0wei160 1 P aa e 161 Page 162 1 P a g e Table S-i: Recognition Category "S" Initiating Condition Matrix (cont.)UNUSUAL EVENTALERTSITE AREAEMERGENCYGENERALEMERGENCYSU6 Loss of a-4ALL onsiteor offsite communicationscapabilities.Op. Modes: P&wei-Pe ... ndk, HaiG sh.fdifw,. ...L2,34SU7 Failure to isolatecontainment or loss ofcontainment pressurecontrol. [42414R]Op. A-r.d.. Po.i. .ot.l..aeiou, saduif,1.. ....1,2.3.4SS8 Loss of al-4ALL VitalDC power for 15 minutesor longer.Op. Modes: P.we;O'c1,2, ',. 5;tz;p, HotSc ...... M,,c, Sh:w:duwnSG8 Loss of a-4ALL ACand Vital DC powersources for 15 minutes orlonger. Op. Modes: Powe*Op12otia.4 , S:or'up, Hct12,_,,4SA9 Hazardous eventaffecting a SAFETYSYSTEM needed for thecurrent operating mode.Op. Modes: Power1,23oio,4 5u p o1631 P a e SulECL: Notification of Unutsual Eveit IUNUSUAL EVENTInitiating Condition: Loss of *IALL offsite AC power capability to emergency buses for 15 minutesor longer.Operating Mode Applicability: Po-ewr 0Oeffltion,, startup. met Standby. t S......O\\ nl1.2, 3,4Example Emergency Action Levels:Note: The Emergency Director should declare the Unusu-al Event, UNUSUAL EVENT promptly upondetermining that 15 minutes has been exceeded, or will likely be exceeded.(1) Loss of A-bbALL offsite AC power capability to (site pethree 4160V AC ESF Busesgasses for 15 minutes or longer.-Mleemet-gency buses).. aIALLBasis:This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plantmore vulnerable to a complete loss of power to AC emergency buses. This condition represents apotential reduction in the level of safety of the plant.For emergency classification purposes, "capability" means that an offsite AC power source(s) is availableto 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 thevia IC SA1.eme..gency ,lassi4iatoin leve .EMERGENCY CLASSIFICATION LEVELwould beSUI: EAL-I Selection Basis:N/AREFERENCES:1. OPOP04-AE-0001, Rev. 44. First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus2. 0POP04-A.E-0004, Rev. 15. Loss of Power to One or More 4.16 KV ESF Bus3. OPSP03-EA-0002. Rev. 32, ESF Power Availability4. Drawing OOOOOEOAAAA, Rev. 24. Single Line Diagram, Main One Line Diagzram. Unit No. I&2Developer- The "site specifie emer-gency are te buses fed by off-ite or emergency AC po'el" s ources that1supply.' 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 beimplemiented within 15 miues onsider capabilitieS suceh HS powerf soHIe ui- ýi~ tie, 'si164 1 P a g e generfatOFS. OtherF poVer SOUrceAS deScribed O in aoral Fr efergenc, beran poedues et.Plantisthat have a poeuazed apability to suipply, offsite AG pEower to aii affected uniit via a el'oss tie to acmpaMIAn fl ay' cr-edit thiS p)OWfr SE*ure in the EAL provided that the planned cr-oss tie strategyEC!I..Assilmelltp,"e'ts of 10 CFR 0A41.1hrRrI -1 1~' 1 A.165 1 P a 14 e SU2ECL: Notiifiation of U.:2H ql :.'2-Event UNUSUAL EVENTInitiating 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-'"4E**mpke-Emergency Action Levels:Note: The Emergency Director should declare the UNUSUAL EVENTUn.sua .Event promptly upondetennining 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 followingparameters in Table S I from within the Control Room for 15 minutes or longer.[BWR paramieter list] [PWAR paramieter list]Reactor Power Reactor PowerRPV Water Level RCS LevelRPV Pressure RGS P .......Primary Containment Pressu.r.e in C)ore/GCre Exit Tem.peratur.eSupprpesi.n Pool Levl Levels in at least (site specifienumber) twe steam generatorsSuppression Poel Temperatur-e Steamn Generator Auixiliary orEmefgeney Feed Water FlowTable SI: Plant Parameters" Reactor Power* RCS Level" RCS Pressure" Core Exit Temperatureo Levels in at least two steamgeneratorso Steam Generator Auxiliary FeedWater FlowBasis:This IC addresses the difficulty associated with monitoring normal plant conditions without the ability toobtain 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 parameterscannot be determined from within the Control Room. This situation would require a loss of all of the166 1P a 2e Control Room sources for the given parameter(s). For example., the reactor power level cannot bedetermined 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 inaccordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC eventreport is required. The event would be reported if it significantly impaired the capability to performemergency assessments. In particular, emergency assessments necessary to implement abnormaloperating procedures, emergency operating procedures, and emergency plan implementing proceduresaddressing 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 ofreactivity control, core cooling [PTJ'] / RPV level [B:jR] and RCS heat removal. The loss of the abilityto determine one or more of these parameters from within the Control Room is considered to be moresignificant than simply a reportable condition. In addition, if all indication sources for one or more of thelisted parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parametersmay 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 SPDSor 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 bevia IC SA2.SIJ2: FAL-1 Selection Basis:The parameters listed were from NEt 99-01. Rev. 6 with the exception of steam generators. Two steamvenerators is a site-specific parameter for the minimum number of steam generators needed for plantcooldown and shutdown.REFERENCES:1. OPOP05-EO-E020. Rev. 11. Faulted Steam Generator Isolation2. OPOPO'-EO-FRHl _ Rev. 23. Response to Loss of Secondary Heat SinkDeveloper- Notes:in tile PAIR paramleter- list columln, the "site specifi nainber'" shauld r-eflect the mininium number ctsteami generaters necessar~y bfr plant eooldown and shutdown. This criter-ionl may, ak'is specifyý whether tileleNvel valuie should be wkide r-ange. niarro-w ranige or lboth. dpnigUpon the moneitoring -r11!e~]inemets illemer~gency oper'ating proce~dmres.Developers may, specify' either pressurizer' er reactor- vessel level in the PW'R parameter' columnn eniti.' AfrR-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 failuremodes,. can hallenge the ability oF an ope..ato ' to accurately' determine, w;ithin the time per.iod m'ai"abfreergeny classificationi assessments, if a specfitc percenftage Of inidicatiOns- haý' beenI je.t. TheapproPach used inl this E .facilitates pr~ompt and accur-ate emei'geney classification assessments by'feetisiing on the insdicationis fori a seleEiea. 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-einiiulmeter- 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 thieassociated ...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 knplementedand 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-rEOP-4. Based on these considefations. a loss of annunciation is conSidfred to be adequiately addr-essed byr~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 10C-FR 5-40.72 (anld associated guiidance in N LREG I 02-2)N. The roting of this event will ensure adequa-teplant staff and NRC an..',.d the establishm.ent of,.. approprFte " ompensa.or. aseandcorrective 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 deSignlbasis efa digital !&C- systemi should conisider, the E-MEA informfatiOn Wilhcn developing thieir site specificDule to cllanges, in tile colnfigurations SAFTY SVSTEMS, inc;luding assocated instrum..entat.ion. and;ndications, duringthe cold ..down.., and o0 analogouS K; is -Fothese moedes of operation+.ECL6 Assignment Attib,'-utesý 3.1. 1A168 1P a e SU3ECL: Notifieation of Unusual Evet UNUSUAL EVENTInitiating Condition: Reactor coolant activity greater than Technical Specification allowable limits.Operating Mode Applicability: PoN.. Opeati n, StAction Levels: (1 or 2)urtup. Hot Standby, Het Shbtdevvin1,2,3.4(1 (St peiiva4e.racilatioii..i......r4RT-8039 readingfeadings greater than 30 uCiicm3(s4ie/Ceelfl-e(2) Sample analysis indicates that a reactor coolant activity value is greater than an allowable limitspecified in Technical Specifications.* Greater than I LCi1-4 4/Pgm Dose Equivalent I- 131* Greater than 100/ EE bar LtCi /gi gross activityBasis:This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in TechnicalSpecifications. This condition is a precursor to a more significant event and represents a potentialdegradation of the level of safety of the plant.Escalation of the emer.gency elas;ifieation level EMEFvia ICs FA 1 or the Recognition Category AR ICs.,GENCY CLASSIFICATION LEVEL would beSU3: EAL-1 Selection Basis:RT-8039 is the Failed Fuel radiation monitor and samples via the CVCS letdown line. [hle value 30LiCi/cm3 is the reading, that is equivalent to I LiCi/gm Dose Equivalent 1-131. The monitor value in thisEAL is the calculated monitor response if the RCS activity were equivalent to I LCi/gln Dose Equivalent1-131. The value is based on Calculation STPNOC013-CALC-003. The value used in this EAL wasconservatively 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 orgreater 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 RiseRCS Activity (RT-8039 EAL Threshold)XXX2. STP Technical Snecification Section 3/4.4.8 Snecific Activity.I 1 ........ ...1691 P a e For EAld1 f[iner tihe radiat ies n mon- i to r(s) tha- ay he utsed to readi rly ideinti f- wh"en RTS acthivlevels exceed Technical Speci'iattion allowable limits. This EAL may be deeloped using diffelxfntm.eh.d .An4d sites q 'hould use ei.tin.g .apabilitieS tO addfeSS 4t (e.g., deVelopMen. .Of nv. ..apab.iliteS iSnot required). Extamples of existing methods...ap.abilities ifude:" An inistalled r-adiation moanitor ein the letdown system or aiv eje faimplenientable conver'sion calculation capability.The mcniater r-eading values should corrFespond to an RCS activity level approxdimately at TechinilSpeiiCatio-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 paaetrsspecified ini Techniical Specifieaiitios anid tie associated allowable limit(s) (e.g, values 1for do~seequiivalent 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.6ECL Assittnmient Aitiribuites: -3.1. k1A and 3. 1. 6B170 Pa, v e SU4ECL: Notification of Unumual Even't UNUSUAL EVENTInitiating Condition: RCS leakage for 15 minutes or longer.Operating Mode Applicability: Power 4)pe:'ation,.S.' Hot Stan.,b., Hotut.own ., 1,2. 3, 4Action Levels: (1 ef-2-or 2 or 3)Note: The Emergency Director should declare the UNUSUAL EVENTUn-usual Event promptly upondetermining 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 15minutes 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 orlonger.Basis:This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCSleakage has been detected and operators, following applicable procedures, have been unable to promptlyisolate the leak. This condition is considered to be a potential degradation of the level of safety of theplant.EAL #1 and EAL #2 are focused on a loss of mass from the RCS due to "unidentified leakage", "pressureboundary leakage" or "identified leakage" (as these leakage types are defined in the plant TechnicalSpecifications). EAL #3 addresses a RCS mass loss caused by an UNISOLABLE leak tgrough aninterfacing 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 ControlRoom indications. Lesser values typically require time-consuming calculations to determine (e.g., a massbalance calculation). EAL #1 uses a lower value that reflects the greater significance of unidentified orpressure boundary leakage..The release of mass from the RCS due to the as-designed/expected operation of a relief valve does notwarrant an emergency classification. Fe- PWRs. a AnaAo emergency classification would be required if amass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticksopen 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 .ndthcrcforc.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-elvia ICs of Recognition Category A-R or F.EMERGENCY CLASSIFICATION LEVEL would beSU4: 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. 6reQluirements 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 throughany one steam generator, but the specification does not specify the type of leakage. Therefore, STPEGSwill use the leakage outside containment: which may include SG Tube Leakage, at 25 gpmi for 15 minutesor 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 thesite'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 thesite'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 SU5ECL: Natifieati,on of Unuusal Evei-t"UNUSUAL EVENTInitiating Condition: Automatic or manual (trip [PArR] 1' ,cram [BWR]) fails to shutdown thereactor.Operating Mode Applicability: PPAWef4 aei--1_ 2Net: A manu.al a.tion is any eperatei action, or. set of actieiis, which cau th.e. ,itf i.d., to bef apidlyMserted int. the coe, and does n .t in.lud. Manually dri 'ing in control rod1 Or .im.plem.ntatin cf bo.roinj..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 berapidly inserted into the core, and does not include manually driving in control rods or implementation ofboron injection strategZies.(1) a. An automatic (trip [P :RI ' sram
- did not shutdown the reactor.ANDb. A subsequent manual action taken at the reactor control eensek--Spanels is successful in shuttingdown the reactor.(2) a. A manual trip (t rp-E ,PR]ANDb. EITHER of the following:rB'WR],rdid not shutdown the reactor.IA subsequent manual action taken at the reactor control ee*+slesgpanels is successful inshutting down the reactor.OR2. A subsequent automatic (trip [1WR] /' [BWR]) is successful in shutting down thereactor.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 takenat the reactor control eonselespanels or an automatic (trip [PWR] / srani [,. -R]) is successful in shutting1731 Pa ye down the reactor. This event is a precursor to a more significant condition and thus represents a potentialdegradation of the level of safety of the plant.Following the failure on an automatic reactor (trip-[P, WR] / scram [BWL R), operators will promptlyinitiate manual actions at the reactor control ce-iseleepanels to shutdown the reactor (e.g., initiate amanual reactor (trip [PWRI / scran [BW-r)). If these manual actions are successful in shutting down thereactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heatremoval systems.If an initial manual reactor (trip [P..R] / scram [9R]) -is unsuccessful, operators will promptly takemanual 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 severalfactors, the initial or subsequent effort to manually ktrip+P-WR4--sc-fa-4BW-R})- the reactor, or aconcurrent 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 Shuttingdown the reactor, core heat generation will quickly fall to a level within the capabilities of the plant'sdecay heat removal systems.A manual action at the reactor control *eonsoelspanels-is any operator action, or set of actions, whichcauses 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 boroninjection strategies. Actions taken at back-panels or other locations within the Control Room, or anylocation 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 varybased upon several factors including the reactor power level prior to the event, availability of thecondenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. Ifsubsequent operator manual actions taken at the reactor control eoeolespanels are also unsuccessful inshutting down the reactor, then the emergency classification level EMERGENCY CLASSIFICATIONLEVEL will escalate to an A4ei4-ALERT via IC SA5. Depending upon the plant response, escalation isalso 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.. RPSsetpoint 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 areapplicable, and should be evaluated.If the signal does not cause a plant transient and the (trip [PWR] / Lcram [ R). faiure isdetenrined through other means (e.g., assessment of test results), then this IC and the EALs are notapplicable and no classification is warranted.SU5: EAL-I, EAIL-2 Selection Basis:N/A1741 P ag e
REFERENCES:
- 1. OPOP03-ZG-0004. Rev. 45. Reactor Startup2. 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 awhi"ch the react.. is sh--utdown. A PWR with a Sh..td.....c....... level that is less, than. orecgual te the rvetacto power- level whichi defines the le,'c boud oIPwer- Operation (Mode 1) will nieed toi:nelude Star'tup (Mode 2) in the Ope.atin.g Mode.Applic.ability. For if the ..eac.tor is, consideredtO be ShutdWownV at Y% and Power- 1 59 Operattioii starts at >50%.. theni the IC is also applicable in Star-tupDevelopesr. maiy include site specific.FOP oriter"ia indicative of. a s.ee.ssf.l r.eactor shut.OW. ini ail-Istatemen"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. maiealtel beaf~s4)EC-L sgnetArius:3L.175 1 P a g- e SU6ECL: Notification of Unusual Event UNUSUAL EVENTInitiating Condition: Loss of aIALL onsite or offsite communications capabilities.Operating Mode Applicability: P..i... Vc....., ht .......... , At i.Wtd.w. 1, 2... 3 41Fimple-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 communicationsmethods 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) lineTable S2: Communications MethodsEAL-1 EAL-2 EA L-3ONSITE ORO NRC* Plant PA svstem Xo Plant Radios X" Plant telephone system X X X* Satellite phones X X* Direct line from Control Rooms to Bay X XCitv* Microwave Lines to Houston X X* Security radio to Matagordca County X* Dedicated Ring-down lines X" ENS line XBasis:This IC addresses a significant loss of on-site or offsite communications capabilities. While not a directchallenge 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 communicationspossible (e.g., use of non-plant, privately owned equipment, relaying of on-site information viaindividuals 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 plantoperations.EAL #2- addresses a total loss of the cominunications methods used to notify all OROs of an emergencydeclaration. The OROs referred to here are (see Developer Notes)Matagorda County Sheriffs Office. andTexas 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 emergencydeclaration.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 relavingof infornation. Links were obtained from procedures OPGP05-ZV-001 I. Elmereencv Communications.REFERENCES:I. OPGP05-ZV-00 1. 1, Emergency CommunicationsDeveloper- Notes:9.. ^ 1 The "site speci..c list of co..u.ni.. tio m.etlhods" .-.uIld ., 11 inOmudo ll LlniEationS mlethodSused 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.. andmaintainied by. indi'.idupls.161 EAL ff2 The "site specifi list of communications methods" should includ!Ee all eommunlicationsm.eth.ds u.sed to pe....rf .. initial emer.enc. n .... --...ati. to ORC. s as described in the site EmerzgencyPlan. The liSting shOu~ld include installed planit equkipmen~t and eomponents. and! not items owned Andmaintained by indiv'iditals. Gsaamplic methods aire indondic tedtlephone lines, cmercain ltelephone 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 eaEAl., f3 The "site speciflc list inicatiOns Shold inlude all .omL i atiOmlethod)ES used to perfOnaR inlitial emer-gency notificatiOns to the NRC as dese-ribed in the site EmergenceyPlan. The listing Shuld.. include ins;talled plant equiipment and components. and not item ; s owned andmainaitainied 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.C1771 Pagee SU7ECL: Notification of Un'usual Event UNUSUAL EVENTInitiating Condition: Failure to isolate containment or loss of containment pressure control.f{4Pw'ROperating Mode Applicability: PVwe pe-iatkS a:upetp, -a-nl-l -utdoewn 1. 2.3, 4Exampie Emergency Action Levels: (1 or 2)(1) a. Failure of containment to isolate when required by an actuation signal.ANDb. ALL required penetrations are not isolatede4Osed within 15 minutes of the actuation signal.(2) a. Containment pressure greater than (54e .pecie .5 , .sig.ANDb. than .ne .ill t"ain of (site p. efi system or equipment) No Containment Sprayeentainmentspra-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) whenrequired by an actuation signal. It also addresses an event that results in high containment pressure with aconcurrent failure of containment pressure control systems. Absent challenges to another fission productbarrier, 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/accidentcondition (e.g., a safety injection or high containment pressure); a failure resulting from testing ormaintenance 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 plantAOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate therequired penetrations, if possible.EAL #2- addresses a condition where containment pressure is greater than the setpoint at whichcontainment energy (heat) removal systems are designed to automatically actuate, and less than one fulltrain of equipment is capable of operating per design. The 15-minute criterion is included to allowoperators time to manually start equipment that may not have automatically started, if possible. Theinability 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 withIC FSI if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission productbarriers.SU7: EAL-1 Selection Basis:N/ASM7: EAL-2 Selection Basis:If containment pressure reaches 9.5 psig, Containment Spray will actuate. If no train of ContainmentSnrav is otneratine ner desicm, the ability to lower containmaent nressure is contnromised. One train ofContainment Spray (Technical Specifications 3/4.6.2) is defined as one containment spray systemcapable 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 Pressure3. Technical Specifications 3/4.6.29. 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 controls'ste'equip:--nt that sh-ould be operatfi-g per design if the conetain-men't pressure actuationsepon i eahed. if desir-ed, specifle eondition ind ications suchl as parameter, 9'lues call abe-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 SA1ECL: Aei4 ALERTInitiating Condition: Loss of a-4ALL but one AC power source to emergency buses for 15 minutes orlonger.Operating Mode Applicability: o r Oper.tion, Sta-rtup. ..ot Standby, Hat S,,.... HtS d 3, 4Fhitntpte-Emnergency Action Levels:Note: The Emergency Director should declare the A4ei:tALERT promptly upon determining that 15minutes has been exceeded, or will likely be exceeded.(1) a. AC power capability to (site specifi. .y buses) ,aALL three 4160V4164 AC ESFBuses is reduced to a single power source for 15 minutes or longer.ANDb. A-NVANY additional single power source failure will result in a loss of aRALL AC power toSAFETY SYSTEMS.Basis:This IC describes a significant degradation of offsite and onsite AC power sources such that anyadditional 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. ThisIC provides an escalation path from IC SU 1.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying requiredpower 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., anonsite 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 emergencybuses 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 bevia 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 Bus3. OPSP03-EA-0002. Rev. 32. ESF Power Availability4. DrawinQ OOOOOEOAAAA. Rev. 24, Sinale Line Diagram. Main One Line Diagram. Unit No. I& 2Developer Notes:For a power source that has nEltiple the ý and.,or Basis section should reflect theminhium 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 thlatSupply ..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 refltheir 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 thaopelatian 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 beipm 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. Planitsthat have a proceduralized capability to suppl-y offsite AC power o -an-a " .... sisECo a may c-redit this, power soue in tile AL provided that th.e plannied cross tip.trmieets the reureet o 10 CER 50.63EG-L Aessizment Attributes:i 3.1 .241811 Pa R e SA2ECL: Alei4 ALERTInitiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer witha significant transient in progress.Operating Mode Applicability: Pwer Opeation, Suita-ip, Hoet Sta-dby. Hot Shutdown-, 2, 3.4E-aniple-Emergency Action Levels:Note: The Emergency Director should declare the Ale4-ALERT promptly upon determining that 15minutes has been exceeded, or will likely be exceeded.(1) a. An UNPLANNED event results in the inability to monitor one or more of the followingparameters 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 PcwcrRPV Water Lev~el RGS Lev~elRPV Pressure RGS PresseuePrimar-y Containment Pressue in Ex:it TemperatureSuppression Peal Leve Levels in at least (site specificniumber-) two steamf generatorsSuppr-ession Pool Temperaturie Steam Generator Auxiliafy--orEmergency Feed Water: Fl&wTable SI: Plant Parameters* Reactor Power" RCS Level* RCS Pressure* Core Exit Temperature* Levels in at least two steamgenerators* Steam Generator Auxiliary FeedWater FlowANDb. 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 load1821Page 0 Reactor seram rBIA ./-W /trip+P-W-j0 ECCS (SI) actuationThr-ma .....v[o-'r- oscillaqtionIs e-a~ter titaIno/ 1",4, lBBasis:This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during atransient 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 thusrepresents 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 parameterscannot be determined from within the Control Room. This situation would require a loss of all of theControl Room sources for the given parameter(s). For example, the reactor power level cannot bedetermined 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 inaccordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC eventreport is required. The event would be reported if it significantly impaired the capability to performemergency assessments. In particular, emergency assessments necessary to implement abnormaloperating procedures, emergency operating procedures, and emergency plan implementing proceduresaddressing 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 ofreactivity control, core cooling LrD fI] ,' RPV level [B"r] and RCS heat removal. The loss of the abilityto determine one or more of these parameters from within the Control Room is considered to be moresignificant than simply a reportable condition. In addition, if all indication sources for one or more of thelisted parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parametersmay 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 SPDSor 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 bevia 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 andshutdown.REFERENCES:I. OPOP05-EO-EO20. Rev. 11, Faulted Steam Generator Isolation2. OPOP05-EO-FRHI. Rev. 23. Response to Loss of Secondary Heat Sink183 P a , e Developer- Notes.!ill tile PWR paramleter list eaol~iln. the "Site specific number', shoukld reflect thle miiuuber ofsteaml genieiataps niecessar-y for plant eooldowN~ anid shuitdown. This critierion mla", alIso specify, whether. tilelevel value should be- -'Ae-r ,n "a'ge or both, d ipm-r tDev'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 specificplant 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 faýilurfemiodes, can challenge the ability of an operator- to accurately deter'mine, within the timie per-iod av.ailablefo;r P~emergecy classifleatiosi assessmenits., if a specific perceentage of indicationis have been lest. Theappr.oach used ini ,his EAL .acilitates prompt and accurate meirgene' ass<:essments byfaeusin'" -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 ControlREooms.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.9the associated guidance in NtUREG 1022). and repor'ed if it sinifi cantly impairs the capability toperform+... emerg.ency assessments. Compensat...' ..easures far a loss of annunleiation can be readilyimplemented and may, include increased monitoing of main conrol boars an.d maore frequent plantroun..ds, by non licensed operators. Their alerting ...function no.twithstand.e. ......an.nun.iators do fat pr'videthe par.a.eter values or- speeific component stats infrm-ation used to operiate the plait, o. processthrough AOPs or EOPs. Based Em these conisider-ations, a loss of alnnunciationi is consider-ed tobadequbately 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--lprocess or effluent monitor values) is to be adequately bounded by the requirements of 10CER 50.72 (and associated ,o 11.. 1e1 ...n.. 1- "NREi 10212.1' The r.epor.ting of this e.enlt will ensure adequateplant staff an.d .RC awareness, iand drive the establis.ent ofappro prate compensat0o:' measur.es andcorrective actions. in addition, a loss of r-adiation monitor;ing data, by itself, is not a pr'ecur-sor to a mor'es: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 adidiations. during the cold shutdown, refueling, and defucled modes, no C- is incuded forthese modes' orfoper-ation.741841 Page SA5ECL: A#ei4-ALERTInitiating Condition: Automatic or manual (trip [PWR] /ý s;eam [.BWR]) fails to Shutdown thereactor, and subsequent manual actions taken at the reactor controleen-eose panelscontrel eonsoare not successful in shutting down the reactor.Operating Mode Applicability: Pewe Opei-at4ionl. 2Note: A manual actioni is any operator action, of set of aetiens, whieh causes the control rads to bcrerapidly inseited into she core, and does not include mfanually dr-ivinig in control rceds or implemenitation oaboron injection strategies.Emergency Action Levels:Note: A manual action is ANY operator action. or set of actions. which causes the control rods to berapidly inserted into the core. and does not include manually driving in control rods or implementation ofboron injection strategies.(1) a. An automatic or manual (trip [PWR] / scram [BWR) did not shutdown thle reactor.ANDb. Manual actions taken at the reactor control eonselespancls are not successful in shutting down thereactor.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 thereactor control ee-sel-,e&panels to shutdown the reactor are also unsuccessful. This condition represents anactual or potential substantial degradation of the level of safety of the plant. An emergency declaration isrequired even if the reactor is subsequently shutdown by an action taken away from the reactor controleensoe-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, whichcauses 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 boroninjection strategies. If this action(s) is unsuccessful, operators would immediately pursue additionalmanual 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 theControl 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.JThe plant response to the failure of an automatic or manual reactor (trip [PWR] 1 scram [14N[trR]) will varybased upon several factors including the reactor power level prior to tile event, availability of the185 Page condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If tilefailure to shutdown the reactor is prolonged enough to cause a challenge to the core cooling j4WR+ /R4Ww.,ater, le.e'il [BWR.] or RCS heat removal safety functions, the emergency ,lassifiation levelEMERGENCY CLASSIFICATION LEVEL will escalate to a Site Area Emnergency SITE AREAEMERGENCY via IC SS5. Depending upon plant responses and symptoms, escalation is also possiblevia 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 inaccordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure atimely emergency declaration.A reactor shutdown is detennined in accordance with applicable Emergency Operating Procedure criteria.SA5: EAL-I Selection Basis:N/AREFERENCES:1. OPOP05-EO-FRSI. Rev. 17, Response to Nuclear Power Generation -ATWSDeveloper- Notes:Thi IC is applicable in any Mode in. whi.h the actual r.eactr. p-wer level ..uld exceed the powe. level a4whici the iS co.Sie... d .ihutd.wn. A IPWR ,ith a r..etor. pOwer; lev.el that is less thani .requal 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 consideredto 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 AsisignmfenlA i,.tlv; ,. i 2 1 '1 i I1861 P ag e SA9ECL: Alet4-ALERTInitiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the currentoperating 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) enHigh winds or tornado strikeF4RE(site specil hazards) Pr-edicted or actual breach of Main Cooling9 Reservoir retaining dike along NorthOther ev.ents with similar hazard char-acteristics as determinied by the Shift ManagefTable S3: Hazardous Events" Seismic event (earthquake)" Internal or external floodiný event" Hi-h winds or tornado strike" FIRE" EXPLOSIONo 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 ManagerANDb. EITHER of the following:1. Event damage has caused indications of degraded performnance in at least one train of aSAFETY SYSTEM needed for the current operating mode.OR2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component orstructure 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 structurecontaining SAFETY SYSTEM components, needed for the current operating mode. This conditionsignificantly reduces the margin to a loss or potential loss of a fission product barrier, and thereforerepresents 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 indicationsfor it will be readily available. The indications of degraded performance should be significant enough tocause 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 orreadily 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 reportinformation. This is intended to be a brief assessment not requiring lengthy analysis or quantification ofthe damage.Escalation of the emefgef'eyvia IC FSI or A-RSIAS4-.elassifieation levcl EMERGENCY CLASSIFICATION LEVEL would beSA9: EAL-1 Selection Basis:The listed hazards are from NEI 99-01 Rev.6 with the exception of the Main Cooling Reservoir breachalon~g the north wall which was included because it is a credible hazard and analyzed in the STPEGSUFSAR.REFERENCES:I. STPEGS U FSAR, Section 3.4.1. Flood ProtectionDcvelopcr Notes:For- (site spceifi hazafds), Edevelepecrs should eansidef ineluding, e~hei sigiiific-ani, site speei fie haixaids tethe- 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 ElECL zinetAri1~ 1 1)D.P188 Pa R e SS1ECL: Site Area Emergency SITE AREA EMERGENCYInitiating Condition: Loss of IIALL offsite and aRALL onsite AC power to emergency buses for15 m11 inutes or longer.I Operating Mode Applicability: P&" ei- Oferation-sta!4up' l= 't S+--- byn 14- Ot ShE!!--1.2. 34I*-v-plEmergency Action Levels:Note: The Emergency Director should declare the Site Area Emergency SITESITE AREAEMERGENCY promptly upon determining that 15 minutes has been exceeded, or will likely beexceeded.(1) Loss of ALL offsite AND ALL onsite AC power to (site specific emergency b ) ALLthree 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 SYSTEMSrequiring electric power including those necessary for emergency core cooling, containment heatremoval/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission productbarrier monitoring capabilities may be degraded under these conditions.This IC represents a condition that involves actual or likely major failures of plant functions needed forthe 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 AllleVel-EMERGENCY CLASSIFICATION LEVEL would beSSI: EAL-1 Selection Criteria:N/ASS4. EA, 1 REFERENCES:I. OPOP04-AE-0001. Rev. 44. First Response to Loss of Any or All 13.8 KV or 4.16 KV BUs2. OPOP04-AE-0004. Rev. 15. Loss of Power to One or More 4.16 KV ESF Bus3 OPSP3-EA-0002, ReN. 32. .ESF Power Availability4. Drawin!g OOOOOEOAAAA. Rev. 24, Single Line Diauram, Main One Line Diagram, Unit No. I1891 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 anAgtecy 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 thatsupply pwErN-@ to the electrical distr-ibutiOnl Sysltemf that poIAer's SAFETY SYSTEMS. There is typicallyemnergency bus per tr-ain of SAFETY SYSTEMS.The EAL and/,or Basis section may specify, use of a nion safety related power source provided taprain 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 sourcesshud 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 beifimpl emented within 15-) minlutes. Con4siderF capabilities such as power source@ cross ties, "swinig"ecaratrsoter--wi' sourcees deseribed in abnormHal or emergenciy ortigrcereset. Plattat 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, acopaio unit na:, credit thiS pow"er source inl thle FAL, provided tat the plannged croass tie strategym~eet: the requirements of 10 CER 50.6-3.OF1-IZLAssignmlenit A0419i1 itt-,; ý -I 1 -4 m1901 P a g e SS5ECL: Site Area Emergency SITE SITE AREA EMERGENCYInitiating 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- _I2EFxampt e Emergency Action Levels:(1) a. An automatic or manual i-trip [PWR] / szcram[BAV]) did not shutdown the reactor.ANDb. AIIALL manual actions to shutdown the reactor have been unsuccessful.ANDc. EITHER of the following conditions exists:..i.. .7t,-4e1 i i c at.. ...f ..n inabilit to ade...atel. ,Cooling -Red entry conditionseenditien metORi~emov3\eheat fr--m the car-e) Core_... ; --pecifie i,.diatin of.an i, abiRed entry conditionseen.d.i.ti... metlity to adequatelyrello','eheat 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 shutdownthe reactor are unsuccessful, and continued power generation is challenging the capability to adequatelyremove heat from the core and/or the RCS. This condition will lead to fuel damage if additionalmitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency SITESITEAREA EMERGENCY.In some instances, the emergency classification resulting from this IC/EAL may be higher than thatresulting from an assessment of the plant responses and symptoms against the Recognition Category FICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additionalthreat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timelydeclaration of a Site Area Emergency SITESITE AREA EMERGENCY in response to prolonged failureto 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 bevia 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 inabilityto 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 toremove heat from the RCS.SS4: EA4,4-REFEIRENCES:1. Procedure 0POP05-EO-F002, Rev. 2. Core Cooling Critical Safety Function Status Tree2. Procedure OPOPO5-EO-F003, Rev. 6. Heat Sink Critical Safety Function Status TreeDeve'cepeo" Notes:This 1C is appi4eab4e in an'; Mode in which the actu-al reactor power level could exeeed the power level atwhichi tile reactor is conisidered shutdown. A PWRwt htonratrpwrlvlta sls hnoequal to dhe rwector poA'.er level whichi defnes the lower bound of'Power Opetation (Mode 1) will need toin.lu.de Sta.tup (Mode 2) in tie Operati. g Mode .A.pplicabi!it'. For ex.ample, if the reactor. is consideredDevclopers miay include site specific BOP cr-iteria inldicativ~e of a suaeeessful reaetor- shutdown in an EALstaten ent. the Basis or both (e.g.. a Ie-a...- powerl.eve4Site 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 vesselwaer, level that dives entry into a cor.e cooling restor.ation proedre (or otherwise ru...e.simplementation of prom..pt r.estorationi actions). Al .at may ise ineore/co'e exit t.erm...upl.etemp.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 tappoxmatlythe middle ofaetive fuwl should use the lowest on scale nidn hat sis n bo.e the topofaetv..e f.el. lithe lowest. scale readin. is abov.. e the top of a.tive fuel, then a ..ea. .v e level val-eshould not he inwacluded.Feorlat that have implemented Westinhous ....e Ownes Group E e.. Respon.se Guidelin.es. enterthe 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 themaain 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 steamgenerators. These parameters shoulId be identical to those used for' the I!adequate Heat Removal thiresholdFutel Clad Barreie Potential Loss 2.13 anid thr-eshold RGCS BarrFier Potential Loss 2.A in the PWIR EALFission ProducIt Barrier Table.ECL- Assignment Attributes: 3.1 ..B-192 P a g e SS8ECL: Site Area Emergency -SITE AREA EMERGENCYInitiating 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 4Ea-mple--Emergency Action Levels:Note: The Emergency Director should declare the Site Area Emergency SITE AREA EMERGENCYpromptly 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 controlSAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plantfunctions 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 bevia 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.3and determined to be 105.5 volts. At 105.5 volts or less, 0POP05-E0-EC0O, Loss of All AC Powerdirects the operators to open the battery output breakers.REFERENCES:1. OPOP05-E0-EC00, Rev. 23. Loss of All AC PowerDpevelprNotes.!The"site specific bus vol.age value" should be based on the .minimum bus v"oltage necessary for adequateoper-ation of 9,AFETY) SYSTF=N4 eq~iipmfent. This v.oltage value sl ld iHCEeFPor-ate a nmaran ofat least 15oinutesooperatioen before thie onset of.. 'bility to operate these leads. This, vltage isusual ylieai-hemimmvakai~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 ,i1app 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 AssIgnmentAitrihltptc' 21 1311941 P a g e SG1ECL: General genc......y GENE.GENERAL EMERGENCYInitiating Condition: Prolonged loss of a-4ALL offsite and a44ALL onsite AC power to emergencybuses.Operating Mode Applicability: Power Operation-, Startup. Hot Standby. V!t Shutdown 1. 2. 3, 4Sa*m+pek-Emergency Action Levels:Note: The Emergency Director should declare the General Emergency GENERALGENERALEMERGENCY promptly upon determining that (hite specific how'6s) hours has been exceeded, or willlikely be exceeded.(1) a. Loss of ALL offsite and ALL onsite AC power to (t"e spec-ific eme...enc.. -c'ALL three4160V AC ESF Buses.ANDb. EITHER of the following:" Restoration of at least one AC emergency 4160VAC ESF bus in less than (site specifiehaEis-)-4 hours is not likely." (Site sPe ifie indiation of all iiability to adeqUat, , remo.ve heat from the core) CoreCooling- Red entry condition metBasis:This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC powercompromises the performance of all SAFETY SYSTEMS requiring electric power including thosenecessary for emergency core cooling, containment heat removal/pressure control, spent fuel heatremoval and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or morefission product barriers. In addition, fission product barrier monitoring capabilities may be degradedunder these conditions.The EAL should require declaration of a Ei,,,g..... GENERALGENERAL EMERGENCYprior to meeting the thresholds for IC FG I. This will allow additional time for implementation of offsiteprotective actions.Escalation of the emergency classification from Site Area Emergency SITESITE AREA EMERGENCYwill occur if it is projected that power cannot be restored to at least one AC emergency bus by the end offo)ur (4) hou~irs-the awflyized stati.n blackotc eid... Beyond this time., plant responses and event1951 Page trajectory are subject to greater uncertainty, and there is an ihefease4higher likelihood of challenges tomultiple fission product barriers.The estimate for restoring at least one emergency bus should be based on a realistic appraisal of thesituation. Mitigation actions with a low probability of success should not be used as a basis for delaying aclassification 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 ACpower 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 entryconditions (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 throughinverters. or by Alternate AC (AAC) sources as defined in NUMARC 87-00. The STPEGS StationBlackout position credits any one. of the three Standby Diesel Generators as the AAC source. Therequired 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 thebasis 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 Bus2. OPOP04-AE-0004. Rev. 15, Loss of Power to One of More 4.16 KV ESF Buses3. OPSPO3-EA-0002. Rev. 32,.ESF Power Availability4. Drawingy OOOOOEOAAAA. Rev. 24. Single Line Diagram. Main One Line Diagram. Unit No. I5. OPO.P05-EO-F002. Rev. 2, Core Cooling Critical Safetx' Function Status Tree6. OPOP05-EO-ECOO, Rev. 23. Loss ofAll AC Power7. STPEGS UFSAR. Section 8.3.4. Station BlackoutDevelopef Notes:_.Althoulgh this WC and EAL may be kiewed as redundant to the Fission Product Barrier- !Gs. it is includedto 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 thatsuppl.'.. power the e@lectical distribution Systen, that pwo'-ers SAFETY SY-STE.MS. There is typically 1emergency 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 statirblackout coping ..nal. ys performed in ac...rdance with 10 ..R, § 50.63 and Reaulator-y Guide 1. 155.stwNiei Bloo6outSit-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 CoolingRPkV Water- Lcvel (as descr-ibed in thle 90P bases).[pJ-] insert' specific A/ exit th1armocouple te-+lpere tur " and/or reactor "essetWater le-el .hat `ri ..ntr into a cor.e .ooling restoatio.in praocedure (or othe-Wise r.equires196 1 P a , e teiiperaturec. grcatei, than 6 200oF and.'E) a reaetor vessel wiater le-vel thaat corrcv~pondc t0 apliromiffatelythle middle ofaetive fuel. Planits withi reactor- vessel level instirurentation 4hat cannot mieasure downl toa t is not aboeactive fuel. if the lowkest ai seale r-eading abv th to faciefel, thn a arieactor vessel level valueTho-ud not be iRCeLIEWed.ForL plantsetat have iniplemcnted Westinghouse Owniiers Gr-oup Emer-geiEYR n Goidelies., enterthe param -t-re -ed in thie C~ore Caookige Red Path.ECL Assi ..... -,,A ibt 3.1.4.9197 1 P a g e SG8ECL: General Emnergency GENERAL EMERGENCYInitiating 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. .. .34Emergency Action Levels:Note: The Emergency Director should declare the General Emer'gency GENERAL EMERGENCYpromptly 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!ALLthree 4160V44460-- AC ESF buses for 15 minutes or longer.ANDb. 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 ACpower compromises the performance of all SAFETY SYSTEMS requiring electric power including thosenecessary for emergency core cooling, containment heat removal/pressure control, spent fuel heatremoval and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor andcontrol SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challengesto 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 atFukushima Daiichi. Minimum voltao.)e for Class 1E 125 VDC battery buses was determined in calculation13-DJ-006 R ev.3 and deternined to be 105.5 volts. At 105.5 volts or less. 0POP05-E0-ECO0, L..oss of AllAC 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 Bus2. OPOP04-AE-0004. Rev. 15. Loss of Power to One of More 4.16 KV LSF Buses3. OPSP03-EA-0002. Rev. 32, ESF Power Availability4. 0POP05-E0-ECOO, Rev. 23, Loss of All AC Power198 1 P a g e
- 5. Drawing OOOOOEOAAAA, Rev. 24. Single Line Diaaram, Main One Line Diagram. Unit No. I&2Developer" Note;:The "site specific emergency buses", ar-e the buses fed by; fiEmor emergency AG powetr sources, thatsupply power' to the electr.ical distribuktiai Sy.stem that power. s SAFETY SYSTEMS. There is tYpic a 1; Iegei- y-bus-,-Per train of SAFETY S-YSP EIThe "site specific voltage ..alue. sh.. ld be based oni the mini.lmum, b.. svoltage necessar. y I:. .adequa.eoperation o ETY SYS TE M eqSipment. This wltage value should incorporate a margin o. at least 15pp~inate-s of operation before the onset of i nabil ity, to. operate those la&.hiS votg sUsually near- theminu4 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, theOeli 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 fbirSAFETY SYSTEMS.This Wc anid EAI.. wer-e added to Revision 6 to aEddi".iatm _Xpeienee from thie M,'afh, 2011aeccident at Fulaushima Daiichi.ECL6 Assiagnmient Attributes: 3.1.4.l3199 1 P a ge APPENDIX A -ACRONYMS AND ABBREVIATIONSAC ................................................................................................................................. Alternating CurrentAOP ........................................................................................................... Abnormal Operating ProcedureAPPAII ............................................................................................................. o,-,erage Power- Range NM eterATW S .............................................................................................. Anticipated Transient W ithout ScramB&W ........................................................................................................................... Babcck and W ilco-lUT ................................................................................................ Beron Injection Initiation Temper-atureBWBR ........................................................................................................................ Boiling B i W ater ReactorCDE ................................................................................................................ Committed Dose EquivalentCFR ................................................................................................................ Code of Federal RegulationsCTM T/CNM T ......................................................................................................................... ContainmentCSF ........................................................................................................................ Critical Safety FunctionCSFST ................................................................................................. Critical Safety Function Status TreeDBA ......................................................................................................................... Design Basis AccidentD C ......................................................................................................................................... D irect C u rrentEAL ...................................................................................................................... Emergency Action LevelECCS ...................................................................................................... Emergency Core Cooling SystemECL ........................................................................................................... Emergency Classification LevelEOF ............................................................................................................ Emergency Operations FacilityEOP .......................................................................................................... Emergency Operating ProcedureEPA ........................................................................................................ Environmental Protection AgencyEPG .......................................................................................................... Emergency Procedure GuidelineEPRG ...........................................................................................Ef....gell. .Pla ime plemienip nge Prucedure91214 ......................................................................................................... Emergency M aagýP w emen t A en yEPRI .................................................................................................... Electc PFi lwer Research IsistitUteERG ............................................................................................ Emergency Response GuidelineFEM A ......................................................................................... Federal Emergency M anagement AgencyFSAR .............................................................................................................. Final Safety Analysis ReportGE ................................................................................. G .. ..l Em .eF, .. ...yGENERAL EM ERGENCYHCFSL ........................................................................................ndependentpHeat CapacityF Temper atu aae imitKeff m .......................................................................................................tive N -eut iron .M u -Altp lica tion aco4S! ......................................................................................................... .... HumaC n Systemo n IfnterafaceIC A ................................................................................................................... .... Initiating ConditionID ................................................................................................................. ........ Inside DiameterIPEEE.............................. individuial Plant Examination of &Eternfal Ev~ents (Genieric Letter- 8.8 20)ISFSI ............................................................... Independent Spent Fuel Storage InstallationKeff ..................................................................... Effective Neutron M ultiplication FactorLCO ...................................................................................................... Limiting Condition of OperationLOCA ................ ....................................................................... L ossR of Coolant AccidentM CR ..............................................................................................................................m ain C-01trol1 ROOMM SIV ................................................................................. M ain Steamn Isolation ValveM SL ........................................................................................................................ Mar Stear m LinemR, mRemn. mrem, rnREM.....................................................mrilli-Roentgen Equivalent ManM W ......................................................................................................... M egawattNEI ........................................................................................................................ Nuclear Energy InstituteNPP..............................2Nuclear Power Plant------- --------- 200 1 P a R e N RC ......................................................................................................... N uclear Regulatory Com m issionN SSS ........................................................................................................... N uclear Steam Supply SystemNORA D ............................................................................ N orth Am erican Aerospace Defense Com m and(N O)UE ................................................................................................... (Notification Of) Unusual EventN UM ARCNU. .A RC.9......................................................... Nuclear M anagem ent and Resources CouncilOBE ................................................................................................................ Operating Basis EarthquakeOCA ....................................................................................................................... Owner Controlled AreaODCM,/ D AN4 ............................................................... Offsite Dose Calculation_ ..ssessm ef.. .-M anualORO ........................................................................................................... O ff-site Response OrganizationPA ........................................................................................................................................ Protected AreaPACS ............................................................................................. Priorit)y Actuation and Control SystemPAG ................................................................................................................ Protective Action GuidelinePICS ............................................................................................ Process inform ation and Contre SystemPRA/PSA ............................................... Probabilistic Risk Assessm ent / Probabilistic Safety AssessmentPW R ................................................................................................................... Pressurized W ater ReactorPSIG ............................................................................................................ Pounds per Square Inch GaugeR .................................................................................................................................................... RoentgenRCC ..................................................................................................................... Reactor Control ConsoeRCIC .......................................................................................................... Reactor Core. l5sol!tion Ced ingRCS ....................................................................................................................... Reactor Coolant SystemRein, remn, REM ................................................................................................. Roentgen Equivalent M anRETS ................................................................................. Radiolo gical Effluent Technlcal SpecificationsRPS .................................................................................................................... Reactor Protection SystemRPV ........................................................................................................................ Reactor Pressure VesselRVW L4.- .................................................................. Reactor Vessel W ater Level Instr.. en. at.. n SystemRW CU .................................................................................................................... Reaector W ater CleanupSAR ......................................................................................................................... Safety Analysis ReportSA9 .......................................................................................................................... ......... Station Blac.l. utSCBA ................................................................................................. Self-Contained Breathing ApparatusSG ..................................................................................................................................... Steam GeneratorS1 ......................................................................................................................................... Safety inlectionSICS Safety inform.ation and SystemSPDS ....................................................................................................... Safety Param eter D isplay SystemSRO ............................................................................................................. ....... .Seni or Rea tor Oper tTEDE ................................................................................................... Total Effective Dose EquivalentTOAF .......................................................................................................................... aTop of Active FuelTSC ............................................................................................................. Technical Support CenterWOG.............................................................................. Westinghouse Owners Group201 Page APPENDIX B -DEFINITIONSThe following definitions are taken from Title 10, Code of Federal Regulations, and related regulatoryguidance documents.A4e4ALERT: Events are in progress or have occurred which involve an actual or potential substantialdegradation of the level of safety of the plant or a security event that involves probable life threateningrisk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases areexpected 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 involveactual or IMMINENT substantial core degradation or melting with potential for loss of containmentintegrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releasescan be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate sitearea.Notificatieti all Un.usual EventUNUSUAL EVENT (-NOU-EIE4-U -0: Events are in progress or haveoccurred which indicate a potential degradation of the level of safety of the plant or indicate a securitythreat to facility protection has been initiated. No releases of radioactive material requiring offsiteresponse 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 involveactual or likely major failures of plant functions needed for protection of the public or HOSTILEACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment thatcould lead to the likely failure of or; 2) that prevent effective access to, equipment needed for theprotection of the public. Any releases are not expected to result in exposure levels which exceed EPAPAG exposure levels beyond the site boundary.The following are key terms necessary for overall understanding the NEI 99 01 emergency classificationscheme.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 givenem'crgeney elassifieati"n le-'LEMERGENCY CLASSIFICATION LEVEL.Emergeny' Classifieati iaLel EMERGENCY CLASSIFICATION LEVEL (ECL): One of a set ofnames or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normalevents or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsiteand offsite response actions. The emergency elassificatieHn IevesEMERGENCY CLASSIFICATIONLEVELS, 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 fouremergency classification levelsEMERGENCY CLASSIFICATION LEVELS by virtue of the potential oractual effects or consequences.Selected terms used in Initiating Condition INITIATING CONDITION and EMERGENCY ACTIONLEVEL Emergency Action LevelEMERGENCY ACTION LEVEL statements are set in all capital letters (e.g., ALL CAPS). These wordsare defined terms that have specific meanings as used in this document. The definitions of these terms areprovided below.CONFINEMENT BOUNDARY: (Inse,. a specific definition for this term.) Dc'vcloper Note Thebarrier(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 NoteThe procedurall d ..efined cond.itionS Or actions taken to secure containment (priimary or sec.ndar. yfor BAIR) and its associated structur~es, systenis, and components, as a functional barrier to fissiopr.od ct .el.aS. under shUtdowN'll c ,nditions. Those actions necessary to place the RCB in theclosed containment condition that provides at least one integral barrier to the release ofradioactive material. Sufficient separation of the containment atmosphere from the outsideenvironment is to be provided such that a barrier to the escape of radioactive material isreasonably 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 threatmessage exists that supports the threat, or (3) A specific known group or organization claimsresponsibility 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 anelectrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically beconsidered an explosion. Such events may require a post-event inspection to determine if the attributes ofan 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 ofsufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to becomecompletely 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 oroverheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOTrequired if large quantities of smoke and heat are observed.HATCH MONITOR: Temporary monitor installed when Containment High Range Radiation MonitorsRT-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 thestation.HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force todestroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includesattack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used todeliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTIONshould not be construed to include acts of civil disobedience or felonious acts that are not part of aconcerted 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 bystealth and deception, equipped with suitable weapons capable of killing, maiming, or causingdestruction.IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relativelyshort period of time regardless of mitigation or corrective actions.INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed andconstructed for the interim storage of spent nuclear fuel and other radioactive materials associated withspent fuel storage.NORMAL LEVELS: As applied to radiological IC/EALs, the highest reading in the past twenty-fourhours 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 theplant Protec-ted Ar-ea perifnetcr (e.g., a site with a large OCA wher-e some porioins of the boundary, maybe a significant distance from; the Protected Ar-ea).. in these cases., develeper-s should consider- using theboundary definied by the Restricted Or Secured Ownier Controlled Area (ROCAiSOCA). The area andboundary slected for shemife Else muitst be eeonsistenit with the description of the sam~e area and boundar-ycont-ained in the Secur-ity Plan.-The area Surrounding thle PROTECTED AREA where SIP NuclearOperating 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 protectiveactions 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 istypically taken to iean the The area under continuous access monitoring and control, and armedprotection as described in the site Security Plan.REFUELING PATHWAY : (Insert a site specific defin.iti.n fc. this term.) Deeloper. Note Thisdes.riptiOn ;incl,, Includes all the cavities, tubes, canals and pools through which irradiated fuelmay be moved, but not including the reactor vessel.RUPTURE(D): The condition of a steam generator in which primary-to-secondary leakage is of sufficientmagnitude 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 itin 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 thatconstitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation tothe 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 securitysystem of reduction of the operational effectiveness of that system. A security event can result in either aSECURITY CONDITION or HOSTILE ACTION.SITE BOUNDARY: The edge of the plant property whose access may be controlled by STPEGS. Thisboundary 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) anexpected plant response to a transient. The cause of the parameter change or event may be known orunknown.THYROID CDE: The dose equivalent to the thyroid firom an intake of radioactive material by anindividual during the 50-year period following the intake.VALID: An indication, report or condition is considered to be VALID when it is verified throughappropriate means such that there is no doubt regarding the indicator's operability, the condition'sexistence, 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 verificationmethods should be completed in a manner the supports timely emergency declaration.VISIBLE DAMAGE: Damage to a component or structure that is readily observable withoutmeasurements, testing, or analysis. The visual impact of the damage is sufficient to cause concernregarding the operability or reliability of the affected component or structure.205 1 P a e e APPENDIX C PERDACANENTLY DEFUELED STATION T/sEATLsReeognitioni Category PD proevides a stand alon~e set 4f I4s/EA6s fat- a Per-manentfly Defuielednuelear pov.'cr planit to consider- for- tis ini dcvelopinge a site spccific emergency, classificatioschemie. For- development, it was assumied that the plant had operatedd unM.d~er a 10 GCFR, § 5 0 lieenfis eand that the operating company has pe.a.nen.tly ceased plan. t oera..tionS. Further, the companyintends 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 theNRC for- exemiptioni froem specific emer-gency platnin requ.ireents. These exNemptions r-eflect thelowered radiological, souce term and risks associated with spent fuel pool storag- e rtelative to r'eactorat powAer operationl. Scuree termiis and accident analyses associated with plausible accidents ar-edocumented in the station's Fina.l Safety Analysis Repo.t (FSAR), as updated. As a result, eachlicenfsee will nieed to develop a site specific emergency classification schemie using the NRappo.e eem.ptions, revised source terms., and revised aeeident analyses as documented in thestation's FSAR.Recognitioni Categtory PD uises the samfe EC~ns as oper-ating reactors; however, the source term andaccident analyses typically limit the EChs to an Unusual Event and Alert. The Unusual Evenit W~sprov idefo an incrieased- awrns f abnormnal conditions while thie Alert l~s are specific to actualor potenitial impacts to spent fuel. The source teigms mnd release motive forces associated with apermanently defueled plant wouild not be suffieient to r-equlire declar-ation of a Site Area Emergenceyor- Gfeneral Emerg~ency.A permaniently defueled stationl is essentially a spent fuel stor-age faeilil' with the spent fuel is storedin a pool of water that se..ves as both a cooling medium (i.e., removal of decay heat) and shield friomdirect radiation. These primaaryj funtitions of the spentskiel storage pool ar-e thle focus of theReognition Categor.y PD Ws and EALs. Radiological effluient , I anid EALs we"re included topoiea basis for- elassifying events that cannot be readily' classified based oni ani obseirvable eventsor planit conditions alone.Appr-opr-iate W~s and EAL-s from Recoagnition Categories A, C, F, H, and S wvere modifiedadincluded in Recognition Catego,')F PD) to address a spectrum of the evenits that may affect a spent fuelpool. The Recognition Category, PD j~s and EALs reflect the relevant guaidance ini Section 3 of thisdocuent. (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 .theNRC 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 MatrijSUAL EVENT ALERTPD AUI Release of gaseous or liquidradioactivity glreater, than 2 times tile (S-itespeeific efflenit release controllingdoetumient) limiits for- 60 minuites or- loniger.Op. 4Aides.- Aola pplibetPD AU2 UNPLANNED rise in plantr-adiationi levels.Op. oIdes o.. IApplkiPD SU4 UNPLANNEDP spent fuel poolOp.. Modf.. : .... A pp÷ PD HII~ Confirmfed SECURITYCOND ITION ar- threat.Op9. Ahodes: o AppeieablePD HU-2 Hazardouis evient aff-ectingSAFETY SYSTEM equipment necessaryforspent fuiel cooling.Opý. Modes:- Not Appike&oblekPD 1MU3 Other" conditions exist whichin the judgmlent of the Emergency Directorwlaffaflt dleclaration of a (NO)UE.Op. Moeeks: AW~f/ IpplicblPD AMI Release of gaseous or liquidradiactiity esuling in k fsite do)segr-eater- than 10 mr-em TEDE or 50 niremthyteid -PDEOp. Modes." N: ApplipcblePD AA2 UNPLANNED rise in plantr-adiation levels that impedes plant accessTruired to maintain spent fitel integrity.O.Modes. No pphbletPD HAI HOSTILE ACTION withinthe OWNER CONTROLLED AREA orairborne at4ack threat within 30 minutes.Op. Mogdes: Nor. 1pp!ieefbk-PD HA3 Other conditions exist whic-hin theEjudgment of the EmergencyDirector warrant eclelar'ation Of an- Alert-.Op. Modees: No XqApj5ie.ThkTable intended(Revis ion 6)IsPtr uise btv L/\L deveieoneis. inclusion in licensee documients is net required. NEI 99 0!ioNleI- 11IN111- i.. ....... l .............. ..PA i r207 Page P)r ATiECWL Of UUSuaIl initiating Condition: Release of gaseous or liquid radioactivity grcater than 2 times the (site speeificeft.ent r.elease .ontrolling do.ument) limits for 60 minu.tes or loger.Operating AModc-i Appli-bility: Applicabl-eExample 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 60mintes 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 relesduration has exceeeded 60 minutes." if the effluent Pawv, past an effluent monitor is known to have stopped due to actionis to isolate therelease 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 bya 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 greaterthan. 2 times the (site specific effluent release controlling document) limits for 60 minutes orIeingeF.This WC addresses a potential decr.ease in the level of safety ofithe plant as indicated by a low l.v.lr-adiological release that exceeds regulatory commitments forf an extenided per-iod off timle (e.g., anunconti-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 tothe 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 radioactiverelease 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 thatcannot be readily or appriopriatel y classified on the basis of plant conditions alone. The inc.lusion of bothplanit condition and radiological effluent EALS mor)e fully addresses the spectrum of possible accideniteents and conditions.Classification based oni effluent monitor- readings assumes that a release path to the enivironfment isesabisihed if the effluent Dlow. n.t ani efftluent maniteir is known to have stonned due to actions to isolatethle release path, then the efflueft 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 fr30 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-beassociated 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 samploeanalyses or en-ir'nmental s-eys, particularl) on ..n.onitored pathways (e.g., spills of radioacti.veliquids 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 TechnicalSpec.i.fi.cations (RETS) Or, fo. plants that have implemented G.en.e.ic. Letter. 89 0.111, fte ffite DaseCalculation 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 methodologyshould 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 effluentpathways that are noet described in the RýETS or ODC-N1213. if incluided. EAL- val uels fo-r these monitorsshould he determiiined usinig the most applicable do~se,/release limits pr-esented iii the RETS or ODCN'1. it isreogizdthalt -A clalcula-ted EAL value may be below v.hat the mon)IitorF caii read; in that ease, the monitordoes not need to be included in the list. Also, some m~onitors may not be go','ri~ed by, TechnicalSpecifications or- other license related r-elated requiirem~ents; ther-e4foe, it is ipratthlat the associatedEAL 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, 2times a release controel limit. The controellinig documenit typically descr-ibes m~ethodologies for deter-miningeffluent r-adiation mAonitor SetpEointS; thteae mlethoedoAlogies should be used to deter-mine EAL values. ineases wher-e a methodology is not adequately defined, developers should deter-mine values consistent witheffluenit 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? tAeldA4;inistra!ive C'on;iioe Sectiieof /the Technicial Spqecifications and, the Reloeentlo qof Procdual Detai4s of RE-TS tor a toese H r eil 111W el rf9rt3 ieeeszy Wiffe i R=i dill12 This incluides consider-ation of the effluent mionitors descr-ibed in the site emer-gency plani seetion(s) whicaddress 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 INPOrelated to emergency response equipment when conisidering the addition of other effluent monitors..... ......................2.... ..........P...N20 9 1P age For EA6 ft Valkies iii this RA L should be 2 times the setpoint establi shed by the radieaetivity dischargepermIit to war-n of a release that is not in compliance with the specified limits. indexing, the value in thisrnanncr enisures eansistcncey between the EA6 and the setpoinit established by a specific dischre permitD..... JEPeh ..OE.ld reSear.h r .adiation monitor design documents or other information souces to ensurfethat 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,,efeature 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 beyondthe operating or display .an.ge of the installed effluent monito.. in those cases, EAL values should bedetermiined 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 monitorreading. 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 includethe 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 licenseesdo noat have this capability. For those that do, the capability may not be within the scope of tho plnTeenneai Specifiations. A li-enisee mnay request to include an PA/\L using real time dose projectosytem 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 licenseesnoet have this capability. For- those that do, these molnitors may not be controlled and maintained to thqesame level as .lan. e.ui.men .. o. within the s.... of the tlait. Tech'nical S ,ecimfiations. in addi.io.,dor;eadings may be influenced by environmental ...ther factrs. A licensee may request to iiusn F er-imeter MOnitoring systemf; approval w.ill be considered on a case by case basis.ielude an EALýECL AssignmentAttr-ibu,-tes .l N"D I 99 01 (Revision 6)1III~III ý ---, ý -') f) I ') C'-,&2101 Pa g e ECL: Notification of Unusual Eventinitiating Condition: UNPLANNED rise in plant rAdiAtion; Ilev~l.Op.r..ting Modec Applicbiliot.. Not ApplicableExample 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).ANDb. UNPLANNED ise in area Fradiation levels as indiated by ANY Of tie fOllwINlg radiationmn: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- overNORMAL 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 inthe ability to conitrol radiation levels within thie plant or- radioactive materials. Either condition is apotential 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 Etadjacent areas that can be deiected by moitrsin ~ tho eloatEions.The eff.ets of planned ev'olutions so-'here the elevated reading is due torea sthat result from plaaned aradioac-tive w.aste materials.otact;ild be considered. Note that EAL #1 is applicable only in casesi U1NPLANNED water lcv~el drop. EAL #2 exceludes radiation levelvities suchi as use of radiogr-aphic sources and movementoEscalation otthe enicr.ency:' classitication level would be via IC PD AA! or PD AA2.2111 P a g e Dc cIopcr Notcs.2Far- EAL #I Site specific indieationis may incelude inistr-umentation values such as .vater- level and arear-adiatiOn H101140F readings, and porsonnel i-epei~s. 4f available, video eameras may allow. fair remoteobservation. Depending on available instrumnentaticn. the declaration may alse be based on indications Etwater 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 withappropriate justification.ECL Assignment Attributes: 3.1. 1.B212 1P a g e P~D suiECL: l~otifieation of Unusual Eventlnitiating COndItion: U N PL0A N VSpent !Uci pool1 te~perature rise.m m 1 AT .A,Cfi-amw M~'odeC Aoinflicmltv: [Not Aesniteigaiele Emcr-gcncy Action Levels-.U[NPLA[N[NE sn~ent +mie seal temioeraturce riSe tO freater thani Wste snceetme F--IBftsS.This WC addr-esses a eondition that is a pr-ewusor-to a mresi event and repr.esents a potentiadegradation ini the level of satety of the plant. Wt uncorrected, tollin flý- tile P88 v1 Wiloc-cur!, anid resulit inialoss 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 alclationsinthe SAR. Typicaly,ý this temfper-ature is 125' to I 50' F. Spent Fuel Pool temaperature is nonmillyvmaintained well below this point thus allowing time to correct the cooling system mlueinpiitclassific-ationi.ECL AssignmentAttribUtes: 3.1 .1 .A N-E1 99 01 (ReiisiF) n 6)ýSIAI '"Q l pI.~ 10 11 C2 Q213 Pati e PD BUIVVI NcWA4tPP16R Ain If 11411qiil ýVPRntiniftitiig JConclitn! Confirmie SELUK1 LY COUNDITIION op 4hreat.Opcrating Mode Applicability: Not ApplieableE 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 (sitespecific 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 miainitaincooling, of spent ftel, and thus r-epresent a potential degr-adation in the level of plant safety. Securityevents 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 PDTimely anid accurate eommnunieatiainc between SceurK' Shift Supervicion and the Controal Roomn-ifol proper classification Of Acu , related event. of these eveiltc will in"itiateaporate 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 SpentFien Storage hncta!!LAtief Seclurity A-agranz].EAL 41 references (cite specific security shift super-vision) because these are thle individuals trainied toconfirmff that a cceuriitv event is occurring Or has occurr~ed. Tr-ainincg oni security event confirimation andclassification 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 inaccor-dance with (cite specific proceLdure)-.EAL #3 ad-d-ressesr thie thratrom the- impact of an aiircraf on the plant. The NRC HeadquarteirsOperations Officeri (HOG0) will commiuniceate to the licensee if thie thr-eat involves ani aircrfaft. The statusA .... .... ..... I ... ..4 X n l 1 1 1 XIDI ... ....... .. ... .. .. .... .... .. ...4 :D .... .... s uperformied in accordanceewith (site c peiimiibim 1tie Li,. a a OR E)i~iiii tL Lii tiii, eat sEmergency plans and implementing procedures are publi. therefo.re, EALSs hould no."tIncrprte Security sensitive inforimationi. This incluides informfationf that mfay' be advantageouis to a214 P a ge p cteniial ad'vcrzary, sueh as thiepanietilafs eoneerning a speeifie tlifeat of threat lacation. Seca 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-tvilta2151 Pa ae Escalation of the em..ergen.y .lassitcati.n level would be via 1C PP H,,AI.DeveloperNtes:The (site specific secewiity sihift supervision) is the title of the en shiftinoivicluai ressenstioe ioilsupfer-VISIon -T ...- onl bnll SeEHFIENY for~e.The (site specific procedure) is the proceduire(s) used by Control Roomf and/or Secur-ity personnel todeter-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 shouldntinoprte Security sensitive infor-mation. This includes informnation that mfay be advantageous to aptential adversa,', su.h as the pa.iculars aonfcerning a specific ti.reat or threat location-. Seceurity1sensitive 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 numberedr-eferences t s elected events descr-ibed in the Security Plan and associated implementin .......esSuch 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-LAssoig~nient Attr-ibutes: 33. 1.1 .A NEI 99 01 (Revosinn 6) Nm--rem:htar 201 CI IGI2161 Page PDIIU)VC-1 N- IItif4P~tifjr,,'l A4: 11n1Ii'il rF.'"Pt1nitrntin~LonditMan p;arou eavent atteetmo jig E Q '%u Y Q- Y'.[M equipment neeess.a~. ta-r sp~ent A iel-7 -................... j .... [ ..........e6811gOprniting Mode Applienbility: Not Applicablele Emergency A.tion Lcvci...The occurr-ence of ANY of the fellowi" Seismic event (earthiquake)" Internal or external flooding event" High winds or tornado strikeEXPLOSION*(site specifie hazar-ds)*.Other events with similar- hazardeng 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-formance0 VISIBLE DAMAGEperfform its de inThis lC addr.esses a hazardo..s event that .auses damage to at least one tr.ain of a SAFETY SYSTEMneeded fer spenit futel coolinig. The damage muibst be of suifficient miagnitude that thie systemi(s) tr-ainicannot, Or potentially cannot, perform its, design function. This condition reduces the margin toa less orpotenitial less of the fuiel clad barriOer. amid therefore r-epresents a potential degr-adatioii of the lev.el of safetyFor EAL Ie ', the Pfist bullet addresses damage to a SAFETYV SYSTEM train that is ini service/oper-ationsince indications foi- it will be readily available.For EAL I., the seiond bullet addresses damage to a SAFETY SYSTEM train that is not in rseR.ice/oeroti oradily appar-ent tkletigli inidications alonie. Operators will1 make this deteffiniationbasd n he totality of avail-able, event and dlamagge r-epor-t infor-mation. This is intended to be abieassessmient 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 theAlert W.S; PD AA , PD AA2.. PD HAl.. r. , of P1A3Devlper-Ntes:For (site specific hazards), dev~elepers should c-onsider inclueding other significant, site specific haz6ardS tOtheQ 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 ofquipmnent, in accordance with site specifie design criteria.EC-L Assignment Attributes: 3.1.1 A anid 3. 1 1C218 Page PPT T-T Ff1 ~J~tific~th~n "f' Jnwmnl F"entinitiating Condition: Otheir eonditionlS exiSt WhiCh ini !he jUdgmienit Of tile E rgil0,ency Dir1ectorI WarrFantdeclaraton of a (NO)UE.Operating Made Applicability! N]ot ApplicableEx.ample Emcr-gcncy Action Levels:i(1) O~th~e conlditiEnS eXiSt which inl the jubdgment of the Emergency Director indicate that events arein rogessorhave oeeurrced whichi indicate a potential degradation of the level Of Safety Of thep~lant Or indicate a security threat to facility pro~tectiOn has been initiated. No r-eleases ofradioac-tive mateia reurnaff-site r-esponse or monitor-ing are expecte-d unless furFtherdegrFadatieon of safety systems occurs.This 4C addr-esses uinanticipated coniditions noat addressed cxiplieitly elsewhiere butt that warranit deelar-ationo-f an emfergency because conditions exist whichi ar-e believed by the Emer-gency Dii retor to fall under thleeimer-gency classification level descr-iption for a NOUE.219 1Pa e PD A1ECL: .4e44initiating Condition: Release Of gaseouS OF liquid radioactivitY resulting in 6ffsite do)Se greater- than 10mrnem~ TEDE of 50 mrcm thyroid GCDEOperating Mode Applicability: Not ApplicableExample 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 timehas 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 duraionhas exceeeded 15 m~inutes.If the effluent flow, past an effluenit monitor- is known to have stopped due to actionis to isolate the releasepath, 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- emqergencyclassification assessments until the results fro.. m a dose assessm.ent using actual meteoerlogy areavailable.(H) Reading oni ANYý of the foljlowing r-adiation moitor48s greater thani the reading- shownA, forF 15Sminutes or longer:(site specific monitor list and threshold values)(2) Dose assessment. ui actual. eteorology indi.ates doses greater than 10 mrem TEDE or 50mrem 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 indoses greater thanl 10 ireina TEDE or- 50 mrem thyroid CDE at or beyond (site specific dosereceptor 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' Ofi llalaionBaftsThis WC addie. e a release of gaseous oe liquid radieativit' that results in projected or actual offsitedese gfeater than or equial to 11% of thie EPA Prot!ectilve ActiOni GUides, (PAGS). it includes bot220 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 a2211 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 thatcanno! be r-eadily of appropr-iately, classified on the basis of plant conditions alone. The inclusion of boethplant 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 wasestablished 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 iestablished. if the effluent flow past an effluent m.nitor is ..own to have stopped due to actions to isolatethe 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 classificationdiver-sity and m~a:, be used to classify ev~ents that wouild not r-eachi the same ECL6 based on plant condlitionsThe ERA PAGs; are expressed ill term-s; of the sum of the effective dose equivalent (E=DE) and ýthecommitted effective dose equvalent (CEDE), or as the thyroid committed dose equivalen"t (CD9). For thepurpose 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.. 1The EPA PAG uidan fe provides fora the use adult thyroid dose cofnversione freeaose ,sot :nix;me stateshave decided to base protective actions on child thyroid CDE. Nuclear power plant !Gs/EFALs need to beconsisitent With thle pro~tective action methodologies employed by the States within their EPZs. hthyroid CDE dose used in the 1C and EA~s should be adjusted as necessary to align With State protectiv~eac-tion decision making rieraThe "site spec-ifi monitor list and thr-eshold values" should be determlined With c-onsideration of thle0 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 thyroidCDE at the "site specific dose r-eceptor point" (consistent with the calculationl miethodologemployed) for one hourj of exposure." Monitor- readings will be calculated u~sing a set of assumed mcteorolegieal data or- atmolsphericidispersion factors; thie data Or factorFs selected for as@ should be the same -As tho A-s-e employed toealculate the monitor readings for 1c; PDAU." The calculation of monitor readings will also) require use of an assumed reles iopic ix; theselected 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 somex-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 51iiii 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 todistinguish between on site and .ff.si.e doses. The sele.ted distance(s) and/r locations should ,elect thecontent of the emergency plani, and the proccdui-al methodology used to determine off-site doses andP~rotective Action Recommendations. The variation in selected dose receptor points mfeans there may besome 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 ensuriethat 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 purgefeature 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 beyonldthe operating or display range of.the installed effluent monitor.. in those cases, AL values should bedetermiined 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. Tiprovi. si notwithstanding. if the estimated/calc-ulated monitor- reading is geate- tlhan approximately110%11 of the highest accurate moniitor readinig, then developers may choose noet to includc the monitor asan 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.enseesdo not have this capability. For those that do, thie eapability, may not be w.ithin the scope of the plantTechniciial Specifications. A licenlsee mlay, request to incelude an EAL usinig r-eal timae dose prt-Eeetiomsytem 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 donot have this capability. For those that do, these mo m t be entrolled and maintained to thesame 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 EALusing 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 AA2maintain spentfitil:m UNPLANNhD rise in plant radiation level thatfte fieb -H4y-imeec-eas Biant access FequIreAOpernting Mode Applincbility: Not Applirabl eExam~ple Emereenev Action Levecls: HI or 2)1) UNPLANNED dose rate greater than 15 mR'hr in ANY of the fo!!lowin.aesreurncontn~ios ocupancy to maintain; controlI of radicactiye maiteial or operation of systenms neededtoa maintain spent fudel integrity:(2) UNPLANNED Area Radiation Monitor readinigs or survey r-esullts inidicate a rise by 100 mnRhrover 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 equipmenitthat must be operated manually or thiat requires local monitoring, in order to miaintain systems needed tomaintain spent fuel integrity. As used here, 'impede' inceludes hinder-ing or initerfer-ing, provided that thein.terference or- delay is sufficient to sign.ificantly threaten n-ecessary plant access. it is this impaired accessthat 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 forexpected 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 erewithout 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-opriatejati~eatienECL Assignment,Ar-iW;p., 4' 1 1 ?C224 1Page PD 14A!nitiating ConEditior: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborneattack threat within 30 minutes.Opcatig Alde Applicability: Nat Applieablef~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 AREAas 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 CINwthnteONRCOTOLDAEor notification of an aircraft attack thrfeat. This event will r-equlire r-apid response and assistance due to thepessipility or mne attacKi progriessing to the PRTPkU 2LL LUAKLA, or The need to prepar-e the plant and starfor a potential aircraft impact.Timfely and accurate communici atiOnHS betweenl Security Shift Superv'ision anld the ConRol1 RcOOesential for proper classificatiein of a securfity related event.Security plans and termiinology are based on. the- guidance provided by NEI 03 12, Tcinplatofor 11Ft'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 staffandimplemienitation of ensite protective measur-es (e.g., evacuation, dispersal or- shelter-ing). The Alrdeelai'ation will also heighten the awareness of Offsite Response Or-ganizations, all owitig, them to bebetter- 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 arenot a HOSTILE ACTION per'petr'ated by a FORCE. Examples include the crash Of a smallairera.., shots Trm .unte.. pnysia ........ , .......... employees. etc. KepOetlig 01 inese types ol eventsis 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 OWNECONTROLLED AREA. This includes any action directed against an !SFSI that is located within theOWANER 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 iswihn30 minutes. The intent of this EAL is to enisure. tha thr -.ea reated notifiationis are made in.atimiely manner- se that plant personnel and ORC~s are in a heightene2251 P ag e state Of readiness. This EAL is met when the throat :'elated information has b'-on validated in accordancewith (site speeific procedure).The_ NýRC Headquarter-s Oper-ations Offeice (HOO) will comamunieate to the licepnse-e if thea threat involvesan aifrcraft. The statuis and siz~e of the plane may be pr-evided by N0%R4lD through the NRC.in some eases, it may niat be readily apparent if ani aircraft impact with~in the OWNER CONTROLLEDARE was intentional (i.e., a HOST-ILE ACTION). it is expected, although net certain, that notificationby an appropriate Federal ageney to the site w-ould clarfy this this case, the .ppropriate federalaeyisintended to be NORAD, F-B!, FAA or NRC. The emerfgency dleclaration, inceluding One basedon 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 notinoprte Security senisitiv~e information. This incluides information that may be ad.anitaggeous to aptniladversary, such as the paticueblars concer~ningg a specific threat or threat oain.Scrtsniieinformation 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 forsup ersin of the on shift security, foree.Emergency plans and implementing pro.edu.es are public documents; therefore. .EA. s shold notincrpoateSecurity' sensitveifomtin This- inclujdes inform14ation that may be advantageouis to a-ptniladversar-y, suceh as the pafticulars concermnin a pcfi hea rthetloain.Scritysensitive 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 numberedr-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 beNNIrded 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 schemedefinition for the OWNER CONTROLLED AREA.ECL= AssignmentI 0 226 1 P a a e PD 4A3initiating Condition! Othe.i- cnditions exist whi.. h in the judgment efthc Emergency Dil'ector warrantdeclar-ation of aniAlert.Operating Mode Applicability: Not ApplieableExample Emecgcncy Acti.n L.v.ls:.0) Other ..ndition.s exist whichi in thejudgment of the E."ergency Director indicate that events areinprgrssorhav~e occurred A~hiek involve an actual or potential sulbstantial djegr:adationH Of tilelevel of safety of the planit Or a Security' event that invle pi-ebable lif~e threatenling risk to sitepersonnel or- dlamage to site equipment because of HOSTILE ACTION. Any r-eleases areexpected 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 declarationof an emergency becauise conditions eist hic-h -Are believed by the Emaergency Director- to fall under theeemer-gency c-lassification level descr-iption for an Alert.227 1 P a P e
Attachment
4STPEGS Emergency Action Level Deviation, Difference and JustificationMatrix -revisions only STPNOCSTPEGS Emergency Action LevelDeviation, Difference and JustificationMatrix Rev. 0NEI 99-01 Rev. 6 ImplementationAPRIO4 STPEGS EAL DEVIATION/DIFFERENCE/JUSTIFICATION MATRIXTABLE OF CONTENTSFISSIO N PRO D U CT BA RRIER ICS/EA LS .........................................................................................................1 STPEGS EAL DEVIATION/DIFFERENCE/JUSTIFICATION MATRIXFISSION PRODUCT BARRIER ICS/EALSThe following section is configured in a manner that is different from theFission Product Barrier Tables in the STPEGS EAL Technical BasesDocument. Where the Technical Bases Document evaluates all three fissionproduct barriers simultaneously for a specific sub-category, this matrixevaluates each fission product barrier individually for all sub-categories.The significance of this fact is that where the fission product barrier table inthe Technical Bases Document moves vertically through the sub-categories,this matrix moves horizontally. STP EAL DEVIATION/DIFFERENCE/J USTIFICATION MATRIXITh rPchniri fnr i nA,, nr POTFNITIA1 I nAz jif ,rtaI trA. Containment radiationmonitor reading greaterthan (site-specificvalue).IMUL tippliLdUleH .Kt Kao iviomiorRT-8050 or RT-8051greater than 40 R/hrOR2. HATCH MONITORgreater than 90 mR/hr10ot ApplicaoleuitterenceLOSS A.- see Ulobal Lomment#9Loss A.1 -Calc STPNOC013-004 Rev. 2 lowered thesetpoint.Loss A.2 -Lowered Loss A.1resulted in lower Loss A.2ORORB. (Site-specific indicationsthat reactor coolantactivity is greater than300 piCi/gm doseequivalent 1-131).3. Sample analysisindicates that reactorcoolant activity isgreater than 300pCi/gm doseequivalent 1-131.DifferenceLoss B.- See Global Comment#9C __________________________ U _________________________ i ____________ I ____________________________Th.of RCS BarrierA. Lontainment radiationmonitor reading greaterthan (site-specific value).Not ApplicableNot ApplicableNot AoolicableDeviationCalculation STPNOC013-004revised February 2015.Reliable radiation monitorreadings not available due tothe effect of TemperatureInduced Current (TIC) and asetpoint value in closeproximity to the backgroundradiation level. STP EAL DEVIATION/D IFFERENCE/JUSTIFICATION MATRIXPnMOqq nf Cnnta~inmant R2rrigarI nlrI-ntlplltii I AncNot ApplicableA. Containment radiationmonitor readinggreater than (site-specific value).Not ApplicableA 1.RCB Rad MonitorRT-8050 or RT-8051greater than 380R/hrOR2. HATCH MONITORgreater than 840mR/hrV i 1I11UA.1 Calc STPNOC013-004Rev. 2 lowered thesetpoint.A.2 Lowered Potential LossA.1 resulted in loweredPotential Loss A.2_______________________ -I- ________________________ 4 ______________________ __________ .L _________________________
Attachment
5Supporting documents CALC. NO. STPNOC013-CALC-002jEN ER Co N CALCULATION COVER SHEET REV. 2EN, Ieodfy.PAGE NO. I of 47Title: Radiological Release Thresholds for Emergency Client: South Texas ProjectAction LevelsProject: STPNOC013Item Cover Sheet ItemsI Does this calculation contain any open assumptions that require confirmation? (if YES, Identifythe assumptions)2 Does this calculation serve as an "Alternate Calculation"? (If YES, Identify the design verifiedcalculation.) Design Verified Calculation No.3 Does this calculation Supersede an existing Calculation? (If YES, identify the supersededcalculation.) 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 ofno 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 inAttachment 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 AreaEmergency and General Emergency thresholds increased from 39.0 to 40.3 and 390 to 403 ptCi/cc, respectively.Study Calculation 1 Final CalculationSafety-Related[ Non-Safety Related[(Print Name and Sign)Originator: Caleb Trainor .- Date: 2/17/2015Reviewer: Chad Cramer Date: 2/17/2015Approver: Date: 2/17/2015Marvin Morris CALC. NO. STPNOCO03-CALC-002tE N E C O N CALCULATIONEveryday REVISION STATUS SHEET REV. 2PAGE NO. 2 of 47CALCULATION REVISION STATUSREVISION DATE DESCRIPTION0 03/03/2014 Initial Issue1 3/14/2014 Resolve inconsistency in decay times for the two calculations.2 2/17/2015 Changes to half-lives in Attachment IPAGE REVISION STATUSPAGE NO. REVISION PAGE NO. REVISION1-9 2ATTACHMENT REVISION STATUSATTACHMENT NO. PAGE NO. REVISION NO. ATTACHMENT NO. PAGE NO. REVISION NO.I 10-22 22 23-29 23 30-47 2 E N E R C 0 N Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002for Emergency Action Levels REX'. 2ExcePpen e-Every project. Every doy PAGE NO. 3 of47Table of Contents1.0 O BJECTIVE/SCO PE .............................................................................................................................. 42.0 SUM M ARY O F RESULTS ..................................................................................................................... 43.0 M ETH O D O F ANA LYSIS ...................................................................................................................... 54.0 INPUTS .................................................................................................................................................... 55.0 REFER ENC ES ........................................................................................................................................ 56.0 ASSUM PTIO NS ..................................................................................................................................... 67.0 STAM PEDE CALC ULATIO NS .................................................................................................... 77.1 Unusual Event -RUI ................................................................................................................................... 77.2 Alert, Site Area and General Emergencies -R4 I, RSI, RG I ................................................................ 8Attachm ent 1 -Hand Calculations .................................................................................................................... 10Attachm ent 2 -Calculations .............................................................................................................................. 23Attachm ent 3 -STAM PEDE O UTPUT ............................................................................................................. 30 CALC. NO. STPNOC013-CA LC-002E N E RC0 N Radiological Release Thresholds REV. 2for Emergency Action Levels REV. 2Excel en e.--,Evety poit Ee',,dy"a PAGE NO. 4 of 471.0 OBJECTIVE/SCOPEThe purpose of this calculation is to determine the Emergency Action Level (EAL) threshold valuesof a radiological release from the Unit Vent or Main Steam Lines for an Unusual Event, Alert, SiteArea Emergency, or General Emergency. The calculated threshold values are to be included in theSTP EAL Technical Basis document, which implements the new NEI 99-01, Revision 6, EmergencyAction Level Scheme and will be submitted to the NRC for approval. Upon NRC approval, the valueswill be used in OERPO1-ZV-IN01, Revision 10, Emergency Classification.Both a hand calculation and the South Texas Assessment Model Projecting Emergency DoseEvaluation (STAMPEDE) software program were used to generate the results. The hand calculationis included as Attachment 1.Revision 1 of this calculation incorporated decay for a release taking place one hour after reactorshutdown. This was done to create continuity between the two methodologies present.2.0 SUMMARY OF RESULTSThe results of the calculations for the radiation monitors specified in the STP EAL Basis Documentand are listed in Table 2.1, below.Table 2.1: Summary of Calculation ResultsEmergency ActionLevelRT-8010B, Unit Vent(pCi/sec)RT-8046 through 8049,Main Steam Lines(AtCi/cc)RU1 Unusual EventHand Calculation 1.40E+05 5.OOE-02STAMPEDE N/A* N/A*RA1 AlertHand Calculation 1.57E+06 4.03E+00STAMPEDE 2.50E+06 4.50E+00RS1 Site Area EmergencyHand Calculation 1.57E+07 4.03E+01STAMPEDE 2.50E+07 4.50E+O1RG1 General EmergencyHand Calculation 1.57E+08 4.03E+02STAMPEDE2.50E+084.50E+02*STAMPEDE was not used to determine the threshold for RU1. Reference 5.10 indicates that the ODCMmethodology should be used to determine the threshold value.This calculation will be used to establish the threshold values for abnormal radiation basedemergencies in the STP EAL Technical Basis document. I CALC. NO. STPNOC013-CALC-002E NERC0 N Radiological Release Thresholds REV. 2for Emergency Action LevelsExcelienre--Every ptoject Every doy f PAGE NO. 5 of473.0 METHOD OF ANALYSISPreviously, STAMPEDE was used to calculate the Emergency Action Level threshold values foreffluent releases. A hand calculation will verify the STAMPEDE calculations. The hand calculation isdescribed in Attachment 1 of this document STAMPEDE conforms to the requirements of STPProcedure OPGP07-ZA-0014, Software Quality Assurance Program. STAMPEDE was run at STP onan STP computer and under the supervision of an ENERCON employee with access to the STP site asa critical worker.4.0 INPUTS4.1 Per NEI 99-01, Revision 6, Initiating condition AUI, EAL 1, the Notice of Unusual Eventinitiating condition is a release of gaseous or liquid radioactivity greater than two times the ODCMlimit 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 ft3/lbm (Reference 5.3).4.6 The release concentration is varied to find the release concentration which correlates to eachemergency action level. Emergency action levels are taken from NEI 99-01, Revision 6 (Reference5.10) for initiating conditions AAI, ASI and AGI. EAL I is the EAL of interest in each initiatingcondition. The doses at the Site Boundary that correlate to the threshold concentrations are listedin Table 4.1.Table 4.1 EAL Offsite Dose Initiating ConditionsAlert Site Area GeneralTEDErem 100 mrem 1000 mremThyroid CDE 50 mrem 500 mrem 5000 mrem5.0 REFERENCES5.1 Offsite Dose Calculation Manual, Revision 17, March 20115.2 Main Steam PORV Capacity Verification MC05591, Revision: 15.3 NIST Steam Tables, 20115.4 OERPO1-ZV-IN01, Emergency Classification Draft Revision 105.5 OERPOI-ZV-TPO1, Offsite Dose Calculations, Revision 215.6 STP Calculation NC-9012, CRMS Rad Monitor Setpoints, Revision 75.7 STP Calculation NC-901 1, Revision 25.8 STAMPEDE Computer Program, Revision 7.0.3.35.9 STAMPEDE Users Manual5.10 NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors5.11 OPGP07-ZA-0014 Quality Assurance Program5.12 ITWMS Call Number 1000010987 Design Document, Revision 0 CALC. NO. STPNOC013-CALC-002Radiological Release Thresholds CAV. Nfor Emergency Action Levels Eee- proe ey Y PAGE NO. 6 of 476.0 ASSUMPTIONS6.1 Unit Vent Noble Gas MonitorTo 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; themonitor reading accurately reflects the Xe-133 release magnitude.To be consistent with ODCM methodology, the main steam line release is assumed to be entirelyXe-133. The noble gas monitor is calibrated to Xe-133 (Reference 5.6).6.2 Release DurationPer Reference 5.10, Sections IC AA1, ASI, and AGI developer notes, the release should beassumed to last one hour.6.3 Release following Reactor ShutdownThe release initiates one hour after reactor shutdown. While a release initiating at reactorshutdown is likely, significant decay of short lived nuclides occurs during the migration time. Arelease at reactor shutdown would have a significantly higher activity at the monitor location thanat the reception site. It is important for the threshold to not be calculated at shutdown as thiswould create a very high threshold which would not be appropriate for releases which occurshortly after shutdown. One hour after reactor shutdown is sufficient time to decay short livednuclides and create a conservative threshold.6.4 Source TermPer Reference 5.1, any unit vent release with increased RCS activity and no core melt should becalculated 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 isappropriate to use an inventory of noble gases plus 0.2% iodine. A steam generator tube rupture isthe only scenario which would create significant offsite doses through a main steam line release.6.5 Default STAMPEDE Input ValuesReference 5.10 developer notes for initiating conditions AA1, AS I and AGI suggest using theODCM or the site's emergency dose assessment methodology. STAMPEDE is used foremergency dose assessment. Per Reference 5.1, when actual meteorology is not available, thedefault STAMPEDE values should be used. Had the ODCM methodology been used, the 500hour peak x/Q value would be used which is less conservative than the X/Q value produced bySTAMPEDE using default meteorological conditions. Therefore, the use of STAMPEDE defaultvalues provides a more conservative estimate than that of the alternative method outlined inReference 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 thesame unit vent x/Q to Units 1 and 2 which would also be applicable to the main steam line. Allreleases are considered to be ground level releases. CA LC. NO. STPNOCOI3-CA LC-002ENER C N Radiological Release Thresholdsfor Emergency Action Levels REV. 2Excee very proec. ,ev doy PAGE NO. 7 of477.0 STAMPEDE CALCULATIONS7.1 Unusual Event -RU I.7.1.1 Unit Vent MonitorAU I recommends declaring an unusual event due to a release of gaseous or liquidradioactivity greater than 2 times the ODCM limits for 60 minutes or longer (Reference5.10).STP sets the ODCM limit at 7.41 E-04 giCi/cc (Reference 5.6, pg. 16). Two times the limitwould be 1.48E-03 ltCi/cc. The threshold is listed in giCi/sec so that variations in flow ratedo 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) 0Equation 7.1.1.17.1.2 Main Steam Line MonitorThe ODCM does not calculate a release corresponding to allowable limits for the mainsteam line monitors. Since the unit vent release calculated in the ODCM was assumed to beprimarily Xe-133, the assumption is made in the ODCM that other noble gases and iodinemay be ignored in the calculation. This assumption is equally justifiable for the main steamline and the same limiting release will be used.The magnitude of the release calculated for the unit vent Unusual Event applies to the mainsteam lines as well. The main steam line PORV's will create a dose exceeding two timesthe ODCM limit by releasing 1.4E+05 ptCi/sec of activity which is equivalent to the releasefrom the unit vent.The steam lines hold saturated steam at 1285 psig, per Reference 5.2, which has a specificvolume of 0.338ft3/Ibm (Reference 5.3). The PORVs will release the steam at 1.05E+06lbm/hr per Reference 5.2. This creates a set flow rate of steam from the main steam lines of2.79E+06 cc/sec as shown below.F
- Density (lbm 28316.846 -ff + 3600 ( e schreC CCm)( th e1.05E" + 06 (-=--/* 0.338 (-- *28316.846 + 3600 2.79E +06-m)' f t3Y hr)secEquation 7.1.2.1 CALC. NO. STPNOCOI3-CALC-002NRadiological Release Thresholds 2for Emergency Action LevelsExcelence--Eve yproject. Ever day PAGE NO. 8 of47Since the flow rate is set, the concentration will determine the limit. Equation 7.1.2.2 solvesfor the limiting concentration of 5.00E-02 /aCi/cc as shown below.Lim iting Release (I --lCI L ( CiLimiting Concentration -Release Rate (-e-).1.40
- 105 (Ci)]1.4
- o~ ~tC) -5.OOE -02 (IICi)2.79
- 106 c(.sec)Equation 7.1.2.27.2 Alert, Site Area and General Emergencies -RA1, RS 1, RG 17.2.1 Unit Vent MonitorInputThe 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 therelease which corresponds to this dose. A release concentration correlating to the EALthreshold value was calculated by varying the input. The following assumptions andinputs 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 valueso Windspeed = 13.2 mpho Stability class DResultsGiven a monitored unit vent release of 2.50E+06 pCi/sec., the Thyroid CDE is 51mrem/hr at the closest portion of the site boundary and the EAL Initiating Condition isexceeded.Threshold values for the Site Area Emergency and General Emergencies are multiples of10 and 100 of the Alert. Since the Correlation between release concentration and dose islinear, threshold values for the steam line monitors are 2.50E+07 and 2.50E+08 ptCi/secfor the SAE and GE respectively. Both are also limited by Thyroid CDE. AdditionalSTAMPEDE 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-002E N E R C O N Radiological Release Thresholds REV. 2for Emergency Action Levels REV.Evefyay. t PAGE NO. 9 of 477.2.2 Main Steam Line MonitorInputA release concentration correlating to the EAL threshold value was calculated by varyingthe input. The following assumptions and inputs were used for this calculation asdescribed 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 valueso Windspeed = 13.2 mpho Stability class DResultsGiven a monitored main steam line release of 4.5 [tCi/cc, the Thyroid CDE is 50 mrem/hrand 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 and100 of the Alert. Since the correlation between release concentration and dose is linear, thresholdvalues for the steam line monitors are 45 and 450 itCi/cc for the SAE and GE respectively. Bothare also limited by Thyroid CDE. Additional STAMPEDE iterations were performed to confirmthis and are attached. Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002ENERCO N for Emergency Action Levels REV. 2Excel!ence--Everyproject Every doy Attachment 1 PAGE NO. 10 of47Attachment 1 -Hand Calculations1.0 OBJECTIVE/SCOPEEach release calculated using STAMPEDE in the main document is calculated by hand in thisattachment and the results compared to STAMPEDE.2.0 SUMMARY OF RESULTSTable 2.1 is displayed again below showing the results from all the calculations. The minor differenceis due to STAMPEDE using decay factors over a one hour period after shutdown. This also accountsfor the change in the limiting dose being TEDE in the hand calculations and Thyroid CDE in theSTAMPEDE calculations. The accuracy of the hand calculation is considered sufficient andrecommended for use in Emergency Action Levels.Table 2.1 ResultsEmergency ActionLevelRT-8010b, Unit Vent(jiCi/sec)RT-8046 through 8049,Main Steam Line(RCi/cc)RU1 Unusual EventHand Calculation 1.40E+05 5.OOE-02STAMPEDE N/A N/ARAI AlertHand Calculation 1.57E+06 4.03E+00STAMPEDE 2.50E+06 4.50E+00RS1 Site Area EmergencyHand Calculation 1.57E+07 4.03E+O1STAMPEDE 2.50E+07 4.50E+O1RG1 General EmergencyHand Calculation 1.57E+08 4.03E+02STAMPEDE2.50E+084.50E+023.0 METHOD OF ANALYSISUsing the limiting dose at the site boundary, the release is back calculated using atmosphericdispersion models. The X/Q value used is calculated from Regulatory Guide 1.145, AtmosphericDispersion Modelsfor Potential Accident Consequence Assessments at Nuclear Power Plants. Ratherthan using the most conservative meteorology, average meteorological conditions are used as inputs Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002EN ER CO N for Emergency Action Levels REV. 2verydy. Attachment 1 PAGE NO. II of 47to most closely agree with STP emergency dose assessment methodology per the ODCM andSTAMPEDE. Assumed nuclide inventories are taken from Reference 5.15. The dose conversionfactors are taken from Reference 5.2. A release concentration is used to find an initial projected doseat the Site Boundary. Using the projected dose at the Site Boundary, the release concentration isscaled 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, March2011 and is 9.44E+07 cc/sec." The main steam line pressure and PORV choke flow rate were taken from Reference 5.5 and are1285 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 is0.338 ft3/lbm.* The release concentration is varied to find the release concentration which correlates to eachemergency action level dose. Emergency action level doses are taken from NEI 99-01 Revision 6for initiating conditions AAI, ASI and AG1. EAL I is the EAL of interest in each initiatingcondition. The limiting doses are listed in Table 4.1. NEI 99-01 Revision 6 states that thesevalues are based on fractions of the Environmental Protection Agencies Protective ActionGuidelines (EPA PAGs) and the General Emergency represents the protective action valuesrecommended by the EPA.Table 4.1 EAL ThresholdsAlert Site Area GeneralTEDE 10 mrem 100 mrem 1000 mremThyroid 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). Thereactor 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 nuclidemixes used in STAMPEDE. The release mixes are a gap release and noble gases plus 0.2% iodinewhich are listed below. Each nuclide mix was normalized to one so it could be scaled to variousrelease activities. Radiological Release Thresholds CALC. NO. STPNOCO 1 3-CALC-002U EN ER CO N for Emergency Action Levels REV. 2Excellence--eyeproject. Everydoay Attachment I PAGE NO. 12 of47Table 4.2 Gap InventoryNuclide1-1311-1321-1331-1341-135Kr-83mKr-85mKr-85Kr-87Kr-88Kr-89Xe-131mXe-133mXe-133Xe-135mMix1.I1OE+051.50E+052.20E+052.40E+052.OOE+051.30E+062.90E+063.70E+055.50E+067.80E+069.50E+061.1OE+056.80E+052.20E+074.20E+06Normalized1. 1 2E-031.53E-032.25E-032.45E-032.05E-031.33E-022.97E-023.78E-035.62E-027.98E-029.72E-021.1 2E-036.95E-032.25E-0 14.30E-02I NuclideXe-135Xe-137Xe-138Cs-134Cs-137Te132Mo99RulO3Ru106Zr95Lai40Ce144Ce-141Sr89Sr9oMix Normalized5.50E+06 5.62E-021.90E+07 1.94E-0I1.80E+07 1.84E-013.70E+/-01 3.78E-072.90E+01 2.97E-074.80E+00 4.91E-081.22E+O1 1.25E-078.80E-03 9.OOE-1 12.90E-03 2.97E- I11.1OE-02 1.12E-101.90E-02 1.94E-107.40E-03 7.57E-1 I1.OOE-02 1.02E-106.40E-02 6.55E-103.20E-03 3.27E-1 1Table 4.3 Noble Gases+0.2% Iodine InventoryNuclide1-1311-1321-1331-1341-135Xe-131mXe-133Xe-133mXe-135Xe-135mXe-137Xe-138Kr-83mKr-85Kr-85mKr-87Kr-88Kr-89Mix6.1OE-028.61 E-021.OOE-011.86E-022.73 E-0 I2.80E+002.40E+024.20E+007.60E+004.OOE-0 11.60E-015.80E-0 I3.70E-017.60E+001.50E+009.80E-012.80E+008.40E-02Normalized2.26E-043.19E-043.72E-046.92E-051.0 1E-031.04E-028.90E-011.56E-022.82E-021.48E-035.93E-042.15E-031.37E-032.82E-025.56E-033.63E-031.04E-023.12E-04(Reference 5.2) are listed in Tables 4.4The dose conversion factors taken from EPA 400R92001and 4.5 below. Radiological Release Thresholds CALC. NO. STPNOCOI3-CALC-002SE ER C O N for Emergency Action Levels REV. 2Ex'ellence--Veryproject Fverydoy. Attachment I PAGE NO. 13 of47Table 4.4 TEDE Dose Conversion FactorsNuclide1-1311-1321-1331-1341-135Kr-83mKr-85mKr-85Kr-87Kr-88Kr-89Xe-131mXe-133mXe-133Xe-135mDose ConversionFactor(rem per uCi*hr/cc)5.30E+044.90E+031.50E+043.1OE+038.1OE+039.30E+011.30E+005.1OE+021.30E+031.20E+034.91.70E+O12.OOE+O 12.50E+02NuclideXe-135Xe-137Xe-138Cs-134Cs-137Tel32Mo99Ru103RulO6Zr95Lal40Ce144Ce-141Sr89Sr90DoseConversionFactor(rem peruCi*hr/cc)1.40E+021.1OE+027.20E+026.30E+044.1OE+041.20E+045.20E+031.30E+045.70E+053.20E+041.1OE+044.50E+051.1OE+045.OOE+041.60E+06Table 4.5 Thyroid CDE Dose Conversion FactorsNuclide1-1321-1331-1341-135Thyroid CDE DCF(rem per uCi*hr/cc)1.30E+067.70E+032.20E+051.30E+033.80E+04The unit vent noble gas monitor energy efficiency by nuclide is taken from Offsite DoseCalculation Manual (Reference 5.3). The values are relative to Xe-133 efficiency since themonitor is calibrated to Xe-133. Table 4.6 displays the energy efficiency by nuclide relative toXe-133. Radiological Release Thresholds CALC. NO. STPNOCOI3-CALC-002E N ER CO N for Emergency Action Levels REV. 2ExceP, p .cvFeyday Attachment I PAGE NO. 14 of47Table 4.6 Energy Efficiency Relative to Xe-133Efficiency Relativeto Xe-133Nuclide (U/Ceuian....................... ....... ..... ............ .(" C ~ s e ~ " eKr-83m *Kr-85m 1.9Kr-85 2.4Kr-87 2.8Kr-88 2.3Kr-89 2.8Xe-131m 0.015Xe-133m 0.14Xe-133 IXe-135m 0.042Xe-135 2.5Xe-137 2.8Xe-138 2.8*There is no relative efficiency available for Kr-83m.omission.t --Assumption 6.4 further justifies theTable 4.7 Nuclide Half LivesNuclide Half Life(hr)1-131 1.93E+021-132 2.38E+001-133 2.03E+011-134 8.77E-011-135 6.61E+00Kr-83m 1.83E+00Kr-85m 4.48E+00Kr-85 9.40E+04Kr-87 1.27E+00Kr-88 2.84E+00Kr-89 5.1OE-02Xe-131m 2.83E+02Xe-133m 5.42E+01Xe-133 1.27E+02Xe-135m 2.60E-01Nuclide Half Life(hr)".Xe-135Xe-137Xe-138Cs-134Cs-137Te132Mo99Ru103Rul06Zr95Lal40Ce144Ce-141Sr89Sr909.08E+006.38E-022.36E-011.80E+042.60E+057.79E+0 I6.62E+0 I9.44E+028.84E+031.55E+034.03E+016.82E+037.77E+021.21 E+032.50E+050 The half-lives are taken from Reference 5.15 which lists the input data used by STAMPEDE. Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002E N E R C O N for Emergency Action Levels RE\'.2Eaceferxp-Every ptojea. Every day Attachment I PAGE NO. 15 of 475.0 REFERENCES5.1 Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident ConsequenceAssessments at Nuclear Power Plants, Revision 1, November 1982.5.2 EPA 400R92001, Manual of Protective Action Guides and Protective actions for NuclearIncidents, 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 75.10 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 Manual5.13 STP Drawing 6C189N5007, General Arrangement Reactor Containment Building, Revision 65.14 NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors5.15 ITWMS Call Number 1000010987 Design Document, Revision 06.0 ASSUMPTIONS6.1 Release lasts for one hourPer NEI 99-01 (Reference 5.14), IC AA1, AS], AGI developer notes,.the release should beassumed 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. Tocalculate the most limiting case, it is assumed that the maximum flow possible is being releasedfriom the PORV.6.2 Nuclide mixPer 0ERP0I-ZV-TPO1, Offsite Dose Calculations (Reference 5.8) any unit vent release withincreased RCS activity and no core melt should be calculated using a gap inventory. It isconservative to assume an increased RCS activity and not within the intended scope of therelevant initiating conditions to assume core melt. Therefore, a gap inventory is used for all unitvent releases.Per 0ERP0I-ZV-TPO1, Offsite Dose Calculations (Reference 5.8) for a main steam line releasefollowing a steam generator tube rupture it is appropriate to use an inventory of 100 percent noblegases plus 0.2 percent iodine. STAMPEDE's source mix for noble gases plus iodine is defined bythe noble gas levels from the coolant mix plus the iodine levels from the coolant mix scaled to apercentage of total noble gas (Reference 5.15). A definition of 0.2 percent iodine means thesummation of iodine levels is equivalent to 0.2 percent of the noble gas levels. Since a steamgenerator tube rupture releasing through the PORVs is the only steam generator tube rupture Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002F4 E N E R C O N for Emergency Action Levels REV. 2Exceflence--Evetyptont. E'ety day Attachment 1 PAGE NO. 16 of47scenario which would create offsite doses large enough to meet or exceed the EALs, thisassumption is made.6.3 Atmospheric DispersionNEI 99-01 (Reference 5.14) developer notes for initiating conditions AA1, ASI and AGI suggestusing 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 defaultSTAMPEDE values should be used. The default STAMPEDE values assume a stability class Dfor atmospheric dispersion and a windspeed of 13.2 mph. These values were used as inputs forthe 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 thesame logic in selecting the appropriate result from the three calculations. The STAMIPEDE valueprinted in the results found in attachment 3 is consistent with the largest of the three handcalculated X/Q values. This suggests that STAMPEDE simply selects the largest of the three X/Qvalues resulting in a much more conservative estimate. This calculation will deviate from therecommendations 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 tobe used. This assumption is also made by STAMPEDE.6.4 Exposure PathwaysThe dose conversion factors used in Tables 4.4 and 4.5 represent a summation of dose conversionfactors for external plume exposure, inhalation from the plume and external exposure fromdeposition. Because the dose estimations are used for implementing early phase protectiveactions, conversion factors using limited pathways are appropriate.The EPA does not provide a dose conversion. factor for Kr-83m. Because the PAGs are based onEPA dose calculations, it is appropriate to only use the nuclides for which dose conversion factorsare provided. Additionally, Kr-83mn represents only 1.33% of the nuclide inventory activity andits exclusion would not significantly affect the final dose.6.5 DecayThe release initiates one hour after reactor shutdown. While a release initiating at reactorshutdown is likely, significant decay of short lived nuclides occurs during the migration time. Arelease at reactor shutdown would have a significantly higher activity at the monitor location thanat the reception site. It is important for the threshold to not be calculated at shutdown as thiswould create a very high threshold which would not be appropriate for releases which occurshortly after shutdown. One hour after reactor shutdown is sufficient time to decay short livednuclides and create a conservative threshold.Decay is incorporated for one hour friom reactor shutdown as well as migration time. Half-livesare taken from Reference 5.15. Migration time is assumed to be the reciprocal of the wind speed. Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002E N E R C O N for Emergency Action Levels REV. 2Exceieernce--veryproject. Fveryday Attachment 1 PAGE NO. 17 of 477.0 HAND CALCULATIONS7.1 Unit Vent Monitor7.1.1 X/QThe atmospheric dispersion factor, X/Q, determines the change in concentration betweenthe unit vent discharge and the dose reception site. This value is based on meteorologicalconditions and will vary with wind speed and stability class. The ODCM uses the highestannual average X/Q value at the site boundary which is 5.3 E-06 sec/r3.However, for anaccident related release STAMPEDE is used rather than the ODCM. STAMPEDE usesreal time, user entered., or default meteorological conditions to calculate the X/Q for aspecific accident. Default values will be used as inputs into the Regulatory Guide 1.145method for calculating X/Q as described below. Default values are identified in section6.0, Atmospheric Dispersion.For a neutral atmospheric stability class, which is the default in STAMPEDE. X/Qvalues can be determined through the following set of equations.X 1Equation 7.1.1.1X 1Q Ulb (3cruycrz)Equation 7.1.1.2X 1Q U10o YuzEquation 7.1.1.3WhereX/Q = relative concentration (sec/m^3)7C =3.14159U10 = 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 1UZ = vertical plume spread (m), a function of atmospheric stability and distance,determined from Regulatory Guide 1.145 Figure 2Vy = (M -1)aysoom + ay = lateral plume spread wvith meander and building wakeeffects (m), a function of atmospheric stability, windspeed U10, and distance; M isdetermined from Regulatory Guide 1.145 Figure 3A = 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-002NE CE 0 for Emergency Action Levels REV. 2Eveirerre-Eeyproyrrt ery day Attachment I PAGE NO. 18 of 47Figure 7.1.1.1: Reactor Building DimensionsEL 241"-0iAssuming the reactor building cross section to be a perfect rectangle and half sphere, thevariables are defined as follows;U10 = 13.2 mph = 5.9 m/suy = 1200 ina, = 4.2 mEy = (M -1)uy8o0m + -y ; M=I --y = 1200 l mA = (135'* 158')+ ( 2)= 31128.37The three equations become;xQ15.9 (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. 2Excelerce-Evety project. Every doy. Attachment 1 PAGE NO. 19 of 47= 1= 3.568*10-6Q 5.9(37r
- 1200
- 4.2)X 1x 1 -1.07
- 10-5Q 5.9
- 7r * [(1 -1)ay800oo + 1200]
- 4.2To select the appropriate X/Q value, the first two X/Q values should be compared and thehigher value selected. This value is then compared with the third X/Q value and the lowerof those two is the appropriate X/Q value. The appropriate X/Q is 5.3 9E-06 sec/m3 fordefault meteorological conditions by the methodology recommended in RegulatoryGuide 1.145.This calculated value is very similar to the ODCM highest average value of 5.3E-06sec/m3 which was not selected for use. Additionally, the value shown in the STAMPEDEoutput file at one mile is 1.032E-05 sec/m3.This suggests that STAMPEDE uses the samemethodology and simply selects the largest atmospheric dispersion value to remainconservative. This methodology will be replicated and 1.07E-05 will be used as the X/Q.7.1.2 Nuclide InventoryAs previously stated, a gap mix is appropriate for this problem. The gap mix is takenfrom 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 emergencyclassifications. The values for the normalized inventory can be found in Table 4.2.7.1.3 Dose Conversion FactorsAs stated in NEI 99-01 (Reference 5.14) developer notes, the purpose of dose projectionsis to check if the Environmental Protection Agencies Protective Action Guidelines (EPAPAGs) have been exceeded. The dose conversion factors provided by the EPA in EPA400R92001 are used. These dose conversion factors account for external plume exposure,inhalation from the plume, and external exposure from deposition and are listed in Tables4.4 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 contributes1.33% of the inventory activity. The lack of this nuclide's contribution to the final dosewill not significantly affect the outcome.7.1.4 Decay TimeOne hour of decay is incorporated to the monitor response due to the release initiatingone hour after reactor shutdown. Decay is also incorporated for the duration of themigration time. The total decay time is one hour plus the reciprocal of wind speed, or1.07575 hours.7.1.5 Dose Calculations Radiological Release Thresholds CALC. NO. STPNOCOI3-CALC-002E N E R C O for Emergency Action Levels REV. EExce.ence--Evev, project. Fvery day. Attachment 1 PAGE NO. 20 of 47The dose rate at the site boundary is calculated using Equation 7.1.5.1.Sn 1.07575/ -F Ci*0.5 T1/2i *DCF1Equation 7.1.5.1WhereD = dose rate per hour at the site boundary-atmospheric dispersion coefficient as calculated in section 7.1.1QF = unit vent flow rateCi= concentration of nuclide i at the time of shutdown1.07575 = the total decay time of interest from section 7.1.4TI/2i = the half-life of nuclide iDCF1 = the dose conversion factor for nuclide i listed in tables 4.4 and4.5The 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 bescaled to find the dose rate of interest.The Alert EAL is 10 mrem TEDE or 50 mrem Thyroid CDE. Using the above method tocalculate TEDE with the appropriate conversion factors, a limiting release rate of2.33E+06 gCi/sec from the unit vent results in 5.7 mrem TEDE. Using the calculatedrelease rate to find Thyroid CDE with the appropriate conversion factors, the samerelease results in a 50 mrem Thyroid CDE at the site boundary. Thus, 2.33E+06 pCi/secis 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 GeneralEmergencies 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 ResponseThe unit vent noble gas monitor is calibrated to Xe-133. Monitor efficiencies relative toXe-133 by nuclide are listed in ODCM Table B3-2. To find the monitor readingassociated with each limiting release, the noble gas concentrations must be multiplied bythe monitor response and summed. Table 4.6 shows the indicated response of the unitvent noble gas monitor by nuclide and Equation 7.1.5.2 shows how the monitor responsewas calculated.
Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002E N E R C O N for Emergency Action Levels REV. 2Excel ence-Every project. very doy Attachment I PAGE NO. 21 of47Monitor Response =
- RejEquation 7.1.5.2WhereC1 = concentration of nuclide i (pCi/cc)Re1 = monitor response to nuclide i (.tCi/cc)x_133 equivalentIn the case of an Alert, the 2.33E+06 ltCi/sec release rate will read as 1.57E+06pCi/sec on the monitor. Kr-83m does not have an indicated monitor responsecoefficient. Because Kr-83m is only 1.33% of the noble gases and does notcontribute to the dose calculation, its exclusion is acceptable.This again is a linear correlation and the SAE and GE scale by factors of 10 and100 respectively.These calculations can be found in Attachment 2.7.2 Main Steam Line Monitors7.2.1 X/QSince the atmospheric dispersion is independent of nuclide inventory or release rate andthe close proximity of thereleases, the X/Q value will be the same for a main steam linerelease as it is for a unit vent release. This assumption is also taken by STAMPEDE andoutline in Assumption 6.3.7.2.2 Nuclide InventoryPer OERPO1-ZV-TPO0, if the release path is the main steam line with a steam generatortube rupture, the nuclide inventory should be 100% noble gas from the reactor coolantand 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 generatortube rupture are taken from Reference 5.15. Values for the reactor coolant inventory arelisted in Table 4.3. All of the noble gases are used and the iodine concentration from thecoolant inventory is scaled to total 0.2% of the total noble gas mix. These inventories arethen normalized to one. These values are listed in Table 4.3.7.2.3 Dose Conversion FactorsThe dose conversion factors used are found in Tables 4.4 and 4.5, taken from Tables 5-1and 5-2 in EPA 400R92001.
Radiological Release Thresholds CALC. NO. STPNOCO13-CALC-002E N E R CO N for Emergency Action Levels REV. 2ExceIence-Evetypfojev. Evetydoy Attachment I PAGE NO. 22 of477.2.4 Decay TimeOne hour of decay is incorporated to the monitor response due to the release initiatingone hour after reactor shutdown. Decay is also incorporated for the duration of themigration time. The total decay time is one hour plus the reciprocal of wind speed, or1.07575 hours.7.2.5 Dose CalculationsEquation 7.1.5.1 applies to the release from the main steam lines. The main steam lineflow rate is used instead of the unit vent flow rate for the value F. The main steam lineflow rate was calculated in Equation 7.1.2.2 of the STAMPEDE CALCULATIONSsection of this document as 2.79E+06 cc/sec.The Alert EAL threshold is 10 mrem TEDE or 50 mrem Thyroid CDE at the siteboundary (Table 4.2). Using the method in Equation 7.1.5.1 to calculate TEDE with theappropriate conversion factors, a concentration at time of shutdown of 4.10 i.tCi/cc wouldresult in 6.89 mrem TEDE at the site boundary if the steam line PORV was open for anhour. Using the same steam line concentration to calculate Thyroid CDE results in 50mrem Thyroid CDE at the, site boundary.The steam line concentrations at the time of shutdown for the Site Area Emergency andGeneral Emergencies are multiples of 10 and 100 of the Alert. Since the correlationbetween release concentration and dose is linear, values for the steam line concentrationat time of shutdown are 41.0 and 410 jiCi/cc for the SAE and GE respectively. Both arealso limited by Thyroid CDE.These calculations can be found in Attachment 2.7.2.6 Monitor ResponseBecause the main steam line monitor is adjacent to the main steam line, significantshielding takes place between the source and monitor. STP calculation NC-90 11 Revision2 calculates a conversion factor for the main steam lines for a noble gas inventory whichis 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 concentrationof 4.10 ltCi/cc at the time of shutdown is 4.03 ptCi/cc. This calculation is also found inAttachment 2. Additionally, the monitor readings for the SAE and GE one hour aftershutdown are 40.3 and 403 piCi/cc respectively. These values are the thresholds for themain steam line monitor. Radiological Release Thresholds CALC. NO. STPNOC013-CALC-0020 E NERCON for Emergency Action Levels REV.2Excelnce-EEvery pq Fery ody Attachment 2 PAGE NO. 23 of 47Table A2-1: Unusual Event Emergency Calculations1.40E+05 I 2.79E+06 I 5.OOE-02Table A2-2: Input Values for Calculations Radiological Release Thresholds CALC. NO. STPNOCOI3-CALC-002I EN ERCO N for Emergency Action Levels REV.2Excelence-Evenyproject Every day Attachment 2 PAGE NO. 24 of 47Table A2-3: Calculations for Boundary Concentrations and TEDE dose due to Unit Vent Release1-1311-1321-1331-1341-135Kr-83mKr-85mKr-85Kr-87Kr-88Kr-89Xe-131mXe-133mXe-133Xe-135mXe-135Xe-137Xe-1381.1OE+051.50E+052.20E+052.40E+052.OOE+051.30E+062.90E+063.70E+055.50E+067.80E+069.50E+061.1OE+056.80E+052.20E+074.20E+065.50E+061.90E+071.80E+071.12E-031.53E-032.25E-032.45E-032.05E-031.33E-022.97E-02..78E-035.62E-027.98E-029.72E-021.12E-036.95E-032.25 E-014.30E-025.62E-021.94E-0 I1.84E-012.76E-053.77E-055.55E-056.04E-055.06E-053..28E-047.33E-049.33)E-051.39E-031.97E-032.40E-032.76E-051.71 E-045.55E-031.06E-031.39E-034.79E-034.54E-031.0 1E-031.0 1E-031.0 1E-031.0IE-031.0 1E-031.0 1E-031.0 1E-031.0 1E-031.0 1E-031.01E-031.0 1E-031.0 1E-031.0 1E-031.0 1E-031.0 1E-031.0 1E-031.0 1E-031.01E-032.79E-08 1.93E+02 2.78E-08 5.30E+043.8 1 E-08 2.38E+00 2.79E-08 4.90E+035.61 E-08 2.03 E+0 I 5.40E-08 1.50E+046.11 E-08 8.77E-0 1 2.61 E-08 3.1 OE+035 11 E-08 6.61 E+00 4.56E-08 8.1 OE+033.3 1E-077.40E-079.42E-081.40E-061.99E-062.42E-062.79E-081.73E-075.61 E-061.07E-061.40E-064.83E-064.59E-061.83E+004.48E+009.40E+041.27E+002.84E+005.1OE-022.83E+025.42E+011.27E+022.60E-0 I9.08E+006.38E-022.36E-0 I2.21 E-076.27E-079.42E-087.79E-071.53E-061.08E-122.78E-081.71 E-075.57E-066.09E-081.29E-064.06E- I11.95E-079.30E+0 I1.30E+005.10E+021.30E+031.20E+032.50E+021.40E+021.1OE+027.20E+025.30E+044.90E+031.50E+041.47E-031.37E-048.11 E-048.09E-053.70E-040.OOE+005.83E-051.22E-073.97E-041.99E-031.30E-091.36E-072.90E-061.11 E-041.52E-051.81E-044.47E-091.40E-04 Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002f ENERCoN for Emergency Action Levels REV. 2EC Nf_47Excellence-Everyproject Every day, Attachment 2 PAGE NO. 25 of 47Cs-134Cs-137Tel32Mo99Ru103Ru 106Zr95Lal40Ce 144Ce- 141Sr89Sr903.70E+Ol 3.78E-07 9.33E-09 1.01E-03 9.42E-12 1.80E+04 9.42E-12 6.30E+04 5.93E-072.90E+O1 2.97E-07 7.33E-09 1.01E-03 7.40E-12 2.60E+05 7.40E-12 4.1OE+04 3.03E-074.80E+00 4.91E-08 1.21E-09 1.01E-03 1.22E-12 7.79E+01 1.21E-12 1.20E+04 1.45E-081.22E+01 1.25E-07 3.08E-09 1.01E-03 3.11E-12 6.62E+01 3.08E-12 5.20E+03 1.60E-088.80E-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 12.90E-031.1OE-021.90E-027.40E-031.00E-026.40E-023.20E-032.97E- II1.12E-101.94E-107.57E- II1.02E-106.55E- 103.27E- I17.33E-132.76E- 124.79E-121.87E-122.52E-121.62E-1 18.07E- 131.0 1E-031.0 1E-031.0 1E-031.0 1E-031.0 1E-031.0 1E-031.0 1E-037.40E-162.79E- 154.83E-151.89E-152.54E-151.63E-148.15E-168.84E+031.55E+034.03E+016.82E+037.77E+021.21 E+032.50E+057.40E-162.79E- 154.75E-151.89E-152.54E-151.63E-148.15E-165.70E+053.20E+041.t0E+044.50E+051.10E+045.OOE+041.60E+064.22E-108.93E-1 I5.22E- II8.49E-102.79E- II8.16E- 101.30E-09Total TEDEDose 5.77E-03 Radiological Release Thresholds CALC. NO. STPNOC013-CALC-00210 E -E RC ON for Emergency Action Levels REV. 2ECOf_47Excellence-E ery proec Ever ydoy Attachment 2 PAGE NO. 26 of 47Table A2-4: Thyroid Dose Calculation for Unit Vent Release1-131 2.78 E-08 1.30E+06 3.61 E-021-132 2.79E-08 7.70E+03 2.15E-041- 133 5.40E-08 2.20E+05 1.19E-021-134 2.61 E-08 1.30E+03 3.39E-051-135 4.56E-08 3 .80E+04 1.73E-03Table A2-5: Unit Vent Monitor Response to Nuclide InventoryKr-83mKr-85mKr-85Kr-87Kr-88Kr-89Xe-131mXe-133mXe-133Xe- 135mXe-135Xe-137Xe-1383.28E-047.33E-049.33E-051.39E-031.97E-032.40E-032.76E-051.71 E-045.55E-031.06E-031.39E-034.79E-034.54E-031.83 E+004.48E+009.40E+041.27E+002.84E+005.1OE-022.83E+025.42E+0 I1.27E+022.60E-0 I9.08E+006.38E-022.36E-012.25E6.28E9.33E8.03E1.54E3.00E2.76E1.69E5.52E7.38E1.28E9.15E2.41E-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 2.8Monitor Reading:0.00E+001.19E-032.24E-042.25E-033.55E-038.40E-094.13E-072.37E-055.52E-033.1OE-063.21 E-032.56E-076.74E-04(uCi/cc) (uCi/sec) Radiological Release Thresholds CALC. NO. STPNOC013-CALC-00210 N ERCO N for Emergency Action Levels REV. 2Excellmen-Everyproject, EverydW Attachment 2 PAGE NO. 27 of 47Table A2-6: Input for Main Steam Line Release CalculationTable A2-7: Calculations for Boundary Concentrations1-1311-1321-1331-1341-135Xe-131mXe-133Xe-133mXe-135Xe-135mXe-137Xe-138Kr-83mKr-85Kr-85mKr-87Kr-88Kr-896.I01E-028.6 IE-021.00E-0 I1.86E-022.73 E-0 I2.80E+002.40E+024.20E+007.60E+004.OOE-0 I1.60E-015.80E-013.70E-0 17.60E+001.50E+009.80E-012.80E+008.40E-022.26E-043.19E-043.72E-046.92E-051.0 1E-031.04E-028.90E-011.56E-022.82E-021.48E-035.93E-042.15E-031.37E-032.82E-025.56E-033.63E-031.04E-023.12E-049.27E-041.3 1E-031.53E-032.84E-044.14E-034.26E-023.65E+006.40E-021.16E-016.07E-032.43E-038.82E-035.62E-031.16E-0I2.28E-021.49E-024.26E-021 .28E-032.9853E-052.9853E-052.9853E-052.9853E-052.9853E-052.9853E-052.9853E-052.9853E-052.9853E-052.9853E-052.9853E-052.9853E-052.9853E-052.9853E-052.9853E-052.9853E-052.9853E-052.9853E-052.77E-083.90E-084.55E-088.47E-091.24E-071.27E-061.09E-041.9 1E-063.45E-061.81E-077.26E-082.63 E-071.68E-073.45 E-066.81 E-074.44E-071.27E-063.82E-081.93E+022.38E+002.03E+O18.77E-016.61E+002.83E+021.27E+025.42E+019.08E+002.60E-016.38E-022.36E-0 I1.83E+009.40E+044.48E+001.27E+002.84E+005.1 OE-022.76E-082.85E-084.39E-083.62E-091.1OE-071.27E-061.08E-041.88E-063.18E-061.03E-086.10E-131.12E-081.12E-073.45E-065.76E-072.47E-079.79E-071.71E-145.30E+044.90E+031.50E+043.1OE+038.1 OE+034.92.OOE+011.70E+011.40E+022.50E+021.40E+027.20E+021.30E+009.30E+015.1OE+02L .30E+031.20E+031.46E-031.40E-046.58E-041.12E-058.95E-046.22E-062.17E-033.20E-054.45 E-042.57E-068.53E- II8.04E-060.OOE+004.49E-065.36E-051.26E-041.27E-032.05E-1 1Total Dose 7.28E-03*Release Constant = X/Q
- duration
- release rate Radiological Release Thresholds CALC. NO. STPNOCOI3-CALC-002EN E RCO N for Emergency Action Levels REV. 2Exctenl --Eer, project Eecry day Attachment 2 PAGE NO. 28 of 47Table A2-8: Main Steam Line Release Thyroid Dose Calculation1-131 2.76E-08 1.30E+06 3.58E-021-132 2.85E-08 7.70E+03 2.20E-041-133 4.39E-08 2.20E+05 9.66E-031-134 3.62E-09 1.30E+03 4.71 E-061-135 1.1OE-07 3.80E+04 4.20E-034.9E0 CALC. NO. STPNOCOI 3-CALC-002Radiological Release ThresholdsFE C O N for Emergency Action Levels REV. 2Excellence-Everyprojct, Every doy Attachment 3 PAGE NO. 29 of 47Table A2-9: Main Steam Line Reading at ReleaseI- I .) 11-1321-1331-1341-135Xe-131mXe-133Xe-133mXe-135Xe-135mXe-137Xe-138Kr-83mKr-85Kr-85mKr-87Kr-88V ,rQO,./ /12-Uq-1.3 1E-031.53E-032.84E-044.14E-034.26E-023.65E+006.40E-021.16E-016.07E-032.43E-038.82E-035.62E-031.16E-012.28E-021.49E-024.26E-021 1)Q 1 l7W22.38E+002.03E+018.77E-016.6 1E+002.83E+021.27E+025.42E+019.08E+002.60E-0 I6.38E-022.36E-011.83E+009.40E+044.48E+001.27E+002.84E+00Iz Inn Al)V./.LiL-U49.77E-041.47E-031.29E-043.73E-034.25E-023.63E+006.3 1E-021.07E-014.22E-044.65E-084.67E-043.85E-031.16E-011.95E-028.62E-033.34E-021.60E-09 CALC. NO. STPNOC013-CALC-002Radiological Release Thresholds REV. 2r E N E R C O N for Emergency Action Levelsknliece--Everyp,,fcFr. Every day. Attachment 3 PAGE NO. 30 of 47DRILL STAMPEDE User Supplied InformationDR LRvso7o.0. t DRELLDautm1ame: 12/1720U 1i524Uasr Nam: Un VeIt AlmT1huS r 11mfiermtAWwlgo Data lupffh:Graund bwd t mud ioty 13M mi~hrOruundLr Itudmdf ID degreesClass$hiy Class: "D -NeuFrnlUaiteedtnit Veal Rinse:kT Veat Release Rate atert 2044E+)6 uCitserReacts St DateTime: 12/1720131424Raluse Start 127172013 15:24EshiaatedRulgase Thonti LOD hmursNuock& 3&tre:.Gsph-teryCakuatedNOBLE GAS relese rate: LDE-0- MuCi/sec.NOmIEASMulisE auI/secNadci nuCtsecPAR37CiEATENMacfi RCisecKr-l3M:Kr-IS:IKr--L.I&Kr47:Kr-U:Xe--In1LMXe-133:Xe-I3M:Xe-135MXe-13TXe-1Im253E-+041.05E+0041.74E+0053.0ff0213.12Em+tIJ9IE+OW41.45E-'0&14E4-003953ECOXD266Ef4041-131:1-132:1-133:1-13:I-135:1112+02/3.IH+06-05B+0153.120+003S.12E+0-3Cs-.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.2.1024E102.SW4D45i3lE-COD1122E4)01352-WDI3.13Eql0112/1712013 3:24:46 PM CALC. NO. STPNOCOI3-CALC-002Radiological Release Thresholds REV. 2E N E R C O N for Emergency Action Levels vbxceflence-Eveyproject. po veryday, Attachment 3 PAGE NO. 31 of47DRILL ~STAA-REDE ResultsIfoatoDRILL R g DRILLlhf&Mr/IWs1172013 15:24 Uer Nan: Unit Vent AlertCnuraammznnflDistauca(miles)051.02.05.07-510.020oDistaxcv(Inits)0.51.02.05.07-510_0Distance(sniles)0-51.02.05.07-510-020.0Ptlmxe Trsrl Tisne( nuns:Msmntas)0:020.050:230-340:451.31CmLQ Va.ln(Sod=*)1.032E-003.755E-05:1.00'E-003.951E-007L541E-0Of7CM BinY243S0053.15EE-OD5737T4-0O?2-441E-0079.10912-00la tz Ikse Dow asIm rsion fl al Body171abrie Z,. 1C~amm0""0.0090.00330:0010:0(X)0.000Isa~rsicn fte Bodynaoil gus gmass (rein)0.0090..001O:0(X)external + anto:0.0160.0020.0010.0000.0w0Imfi a. CUErut (rusr"0.1370-0510-018WXM0.11120.0010.000'IMIE Idize CDEexernsl + internal Muraiu(rem) (rem)0.0160.0060.0M20.0010.0000.0000.0000.1370-0510.0180:0040.002-O.OO20.0010-M012317/2013 3 .2428 PM CALC. NO. STPNOC013-CALC-002for Emergency Action Levels REV. Radiological Release ThresholdsExcellence--Every proect. Every doy. Attachment 3 PAGE NO. 32 of 47STAM,,PEDE Results InformationDRILL R-iina 7.0.3.3 9%W82011 Page2of2 DRILL]CalcuLation ComphtedtRESULTSiMe&odofProjecdon: WindVelocity: 132nIr Release Rate: 119E+006 uCVsecSTAMPEDEWindDirection: ISOOffMie Dose Projection (rem):1 mile 2 mile.? S, miles' 10 milesTEDE 0.006 0.01)02 001 CDE 0.051 0).018 0.004 0.001Projected duration ofreleme: 1.0 hoursA General Emergency Requires a Protective Action RecommendationEVACUATE ZONE(S): ISHELTER LN PLACE ZONE(S): 2AFFECTED DOWeNWIND SECTORS: R. A, BAll Remainine Zones Go Indoors And Monitor EAS Radio StationBased on a Dose Rate Projection of> 3 mreminir (Immersion Ibole Body Noble Gas Gamma) at theSite 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 Director12117!2013 3:24:44 PMDatT'TimeDareTime12117P-013 3:24-.2S PM CALC. NO. STPNOCOI3-CALC-002Radiological Release ThresholdsA E N E R C O N for Emergency Action Levels REV. 2Excei'lence--Every Everyday. Attachment 3 PAGE NO. 33 of47DRILL STAMPEDE User Supplied InformationDa4fr/Tn: 12'1&2-13 07:54Cammnts:9sr Name: Stesminti Site AltIUT SiDkwbrfata Input:hndleel ,Mnd'elofltr 132 mil/hrGroumd "ad dfrrn: 130 d&greestker-select-tedShlYl classStuImtyC is: "D- &bnitwed S)GTnube1pure RIdease:Stea= Adt t5E+000 uC1i/ccStezmT1hwlate: L050 MlhbrRadacr Shm DalteTme-:Rinms Start DatTimn:FklmatedRdeas Duration:12/1312f13 06:54n1mn21on1 07:5410D hearsNudi& Matrme: Mtle Gas +Icimelon. am mpm-atf of gz& 02%CakrlatedNOBLE GAS rn rate: 1gE+07 wCi/secNOELECAS IODIMEPAYCIITIIUEMaCB& uCi/secMucUSl umCI/secKr-UM:1r47:Xe-13&Xe-MmlXe-IM:xe-13atXe4137-Xe-lit:1.1,1E-t-(3.43E+00X55-7E40004255NE+0044.2E-00431.0lM-t0D73.IZENXIS123E40+M127E&101I3ffiXM1-131: 31)E+0031-132: 3.22E+0031-133: 42824GB1-134: 41229+0021-135: L23E+00tCs-li:Cs-137:ce/r-13'/:Ce-141:1.a-lAO:3&t-99:RaIRb-106:Sr/Y-90.Sr-":Te-412:Zr-95:GOJE4WO0.0aE24000GOIXJ-0000.00E+000O.OaE+COOO.XE+tX0O0.3m+10000311+030121M 13 7?55:19 AM CALC. NO. STPNOC013-CALC-002Radiological Release Thresholds REV.2E N E CO N for Emergency Action Levelscxceldence--e'y projwer. day. Attachment 3 PAGE NO. 34 of 47DRILL STAMPEDE Results InformationDP,07033 111 I g laf2 DR ILl "mrzEn 12/1, 20130754 lw Nam: StemI.,m Aatcammmmis051.02.05.0751D.02U.Duuance0.51.02.05.0731.02.05.07.510.020.0mam0:02&230-340:45131CW]/Q Vain.3.750-406T1.00_-00v"5.'D42-0011.5412407C D2 DMI2-43S0059.11aE-0067-373E-0972_441E-0W79.10-S-kkzs~uraie frost RufusIamerniaa Wfmlel Bmdynoble go gamma0.0110o0420.0020.000o.oooomoo0.000anoble em gamma (reim)0.0110.0040.002-0.0000.000IPAGDos RaisIodine CDEexternal +.iflerual MUyrei(--Ifb) (--11r)0.019 0.1350.003 0.0170001 0.0170.001 0.(040.000 0.000 0.0010.000 0.000IME Iodine CDE(lerui+ imainual '&Tria0.0190.0070.0010.0000.0000.1350.0500.0170.0040.0020.0010&012A &'2013 754:42 AM CALC. NO. STPNOCOI3-CA LC-002Radiological Release Thresholdst3 EN ER CO N for Emergency Action Levels REV.&~celence--Everyptoje.r, 'vetydoiy Attachmnent 3 PAGE NO. 35 of 47DRILL PaTA , Results IoraaionILLDR LLRE%'iion 7.0.3.3 9,29,72011 Page2of2DR LMethodof Projection:STA:MPEDEOffsde Dose Projection (rem):I mileTEDE 0.007CDE 0_050l C.acultiom CompletedRESULTSWind Vlocity: 13_2nihrWind Direction: 180Release Rate: l.19E+07uCiisec10 miles0-0012 mifles0.0030.017miles0.001OI0wProjected dsn-afion of'releae: 1.0 hoursA General Emergency Requires a Protective Action RecommendationEVACUATE ZONE(S): 1SHELTER IN PLACE ZONE(S): 2AFFECTED DOWNWIND SECTORS: R, A, BAll Remaminin Zones Go Indoors And Monitor EAS Radio StationBased on a Dose Rate Projection of > 3 mremnir (Immersion Whole Body Noble Gas Gamma) at theSite Botmdary (1 Mile) for 15 minutes or longer the Emergency Classification Initiating Condition RAI(ALERT) has been met.PERFORMED BY:'NameREKIIIMID BY:Rdls'd. nage r:Ramdolo gic al Director121&92013 7:55:14AMDasThne121212013 7:54:42, AM CALC. NO. STPNOCOI3-CALC-002Radiological Release ThresholdsREV. 2E N E R C O N for Emergency Action Levels _EV. 2Exce ence--ver eptoe. Every day. Attachment 3 PAGE NO. 36 of 47DRILL STAMPEDE User Supplied Information,71.033 9/O DRELLDthfrrie: 12/17/201315:25commmtsUser Nam*: UEit Mew Site AmaIar S- s, mnztk Ilh Impxf:Qomn1k-dfl tadikctf 132 nt&rQonmile~wi dfrsrr 1la &grfeeslker-siec-tedStzliiy ClassSthiy (Clas: "D -Neufr."niirmpedklf" Vat 1Rine:Ilit Veat Rlase Rate mteretd 2Z--1007 Reatdr Shutdm Dahtsrne:Rinse Start DatiTume:dRelnseflnraliowNni& Iftdure:12117fD13 14-:251217/2013 15:25100 hemsQzpbnnturyCakuatedNOBLE GAS rmnse rate: 1!91-7i7u -eNOBLEGASN Iadi Cv1sacNadide RCi/sA,MucS& mCilsec10"-80_K,,r-w9:Xe-.131Xe-13iXL-135MXe-137-2527E4051.fl7.06E+M]9.0AE.4X15I:73E+ODS3.14E-03DX3.121+0X14621+-W0191E'0051+J-NX149.74E+0X1-131:1-132:1-133:I-lM:--135:1-135:301SE--N-5.12E+004Cs-1i:Cs-137:Cemr-144:Ce-Idl:Rn/Rh-l06tSr/Y-w-Sr-":Te-132:Zr-0:1.05M+01924E+0002.GE-+O02.94M.3531rE-033.+43o000i24E,-OD425GE40031XIG-OO21.82Edl)135E+tOD3.LV3-M1247T2013 3"2533 PM CALC. NO. STPNOC013-CALC-002Radiological Release Thresholdsr E N E R C 0 N for Emergency Action Levels REV.2.-, y; day. Attachment 3 PAGE NO. 37 of47DRILL STAMPEDE Results InformationDRILL Rv=03%M11 Page In!2 DRILLiDak(Tz 12fl.,2D13 1525 w N'ame: Util Vnt Site AraIDihnaz(miles)0.2.05.07-510.020.0Diltauc,(Mmiks)1.02.007.5IDG2.0L5.07.5l.o20.0Pinm. "wel TMme CELDQ Valn.@ounmiamos)(soc/a')0012 2.&,.0050105 L0372E-000:09 375SE-00-23 1.01D-0060-34 5. 7E-0070:.45 131 1541E-0072-43dE-M09.11CE-003.151E-OW77338-007(3.2458-002441E-09.109E8-EImmursioa fob Bodyxale Zu gunoan(r.am~r)0.0820-0320-0120.0130.0020-aolImmersioaInoal BodyaRti. zgofmm. (rem)0.0120:0030.0020-0010ACO0IPAGDosoeHaloMTE Iodine CME(ron/ic~) (rom/ir)0.160 13640.060 05100.021 0.1760.005 01410.003 0.0210.,00 0.0140.0101 0.005Iodine CDEoxteroul + intern!j MUTSUi0.1600.06D0.0210.0050.0020.0(01136405100-1760.0410.0210.0140M0051217/2D_13 3:25:21 PS CALC. NO. STPNOCOI3-CAL-C-002Radiological Release Thresholds CALC. NO.2SPNOC13-CAC-00EN ER CO N for Emergency Action Levels 1EV.2Fx,, cP-11enrydojy. Evetyda. Attachment 3 PAGE NO. 38 of47DRILLResults information9)2ge2011 Page2of2Ca_.-intions- CompetedRESULTSDRILLleledodofProjectioa: WindVelocity: 132 mi/hr Relee Rate: l.19E-'07 uiQSTA-MPEDEWindflirerdon: 180Of0ste Dose Projection (rem):I mile 2 mile- Pmils 10 mile!TEDE 0.060 0.021 0.005 0.002CDE 0.510 0176 0.41 0.014Projected duration ofrelease: 1.0 houTsA General Emergency Requires a Protective Action RecommendationEVACUATE ZONE(S): 1SHELTER IN PLACE ZONE(S): 2AFFECTED DOWNWTIND SECTORS: R, A, BAll Remaining Zones Go Indoors And Monitor EAS Radio StationnecBased on a Site Boundary (1 Mile) Dose Projection > 0.1 rem TEDE andlor 0.5 rem Thyroid CDE theEmergency Classification Initiating Condition RS 1 (SITE AREA EMERGENCY) has been met. IPERFORMED BY-NameREVIEWED BY:Rod Mhnager:Xnadiological Director12/17/2013 3 25:28 PMDatv'TfieDat&eTime12'17fl013 3:25:21 PIM CALC. NO. STPNOC03-CALC-002Radiological Release ThresholdsREV. 2, E N E R C O N for Emergency Action LevelsExceilence-Everyptojecr. Everyday, Attachment 3 PAGE NO. 39 of 47STAMPEDE User Supplied InformatiOn DRILLDRILL DRism7.033 9M01IDhf/imm: 12f17M2._3 1513CeomatmSlr NamE: StEM LI AXeaIThaf SumuSed]Oeqmdlewi adiwlocity Ifl nil/hGrmmdie tmdsfdrkn 110 LgreesCsmStbbmT Ycl : "D -Neutarxmifsklmed SGTWIhRpInre ReiesSteam Arftif A45RE+001 uoCt/cStemmlhwRzte: LownlEbrRector Shut&=e m DateTime: 12/171913 1413Rase Start DatI/Tume: 12/17/201315:28hiuntedRleas Dwatori LOG homrsNndi& MAfhre: NuMo Gas +LItmeJ&O.. 2S p~rC4t oftmold gas: 01%CakuktINOBL'E GAS ralas rat.: 1102-001 m /secNOHILEGASNmUtb& aG/seCIODINENudkh o42i/sPARflCELIEMNumcis .0/secKr-83M:Kr47:Kr-97-X-13DIMXL-133MXL-lZM:XL-Un:Xe-137-XeL-U381.14E+003.4.5-+00S1IE+0052.55E+00591E7+005127E+0063.SE$XJO6121E-004114E40041]-131:1-132:I-lM21-133:1.-135:323E+0OG4-9aE400$1-24E4005Cs-134:Cs-UL7-Ce/Pr-lit:Ce-141:I.e-lO:lb-l(:RR-l03:Sr/Y9tSr-49:Te-OS:Zir-9:O.O+W00O.C(E+(O0O.OA+ODOO.O[E+O0000.3E+0000.0M+000O.XE+O000.O4IE-tO0O.DXI+OOD12J/7(/013 329:03 PM CALC. NO. STPNOC013-CALC-002Radiological Release Thresholds REV.2F ,.E E C Nfor Emergency Action LevelsEyctllence-Everypo]im. Evetyday Attachment 3 PAGE NO. 40 of 47DRILL STAMPEDE Results InformatoD-RILL Psgelaf2 DRILLIlairm 12117.,2013 15:28 Thu Na: StamLnM SkB ArmCUmmEts:mf15..usmiane=hJt I(miles)031.02.05b07-510-020-0Distance(mniles)0.52.05.07510-02O00DiSLfJe(Mniks)031.02.05.07520.0Plume T~rral Tuao0:050230-340-45[31Value21696-001.032E-003.45E_-0071.0.,'1.-0O63.9513-42US4IE-007CHVQ DflL(cecj')2-4360059.11C.-0067373E4-02-441E-1079-109E-.0I& M .Doe RwIsa~inn Whle Body0.1110-04230.0150.004O-O20.00101131Imaniaz Wfole Bodyoatblh Mgs gamm (rem)0.1110.0420.0150.0040.0020.0010.0017Mlorterusi +.aturnl0.1w90.0720.0250M60.0020.001LODose RatesIodine CD]13540.5060.1750.(04101Ml0.0130.0055 !E Iodize CDEextrnal + inieoal Tkyraid(cern) (Mm)0.189 13540.072 03500.025 0.1750.06 0.0410003B 0.0210.002 0-0130.001 01M0512117i2..013 23".53 PM CALC: NO. STPNOC013-CALC-002SRadiological Release ThresholdsjE N E R C 0 N for Emergency Action Levels REV. 2Excelle,,ce--Everyproect., ve.,day. Attachmnent 3 PAGE NO. 41 of 47STAMNIPEDE Results InformationDRILL id~onTO3 3 9!2&'2011 Page2of 2 DRILLCa-u'donCti ompletedRESULTSWind Velocit- Wind Dir eion: IgoRelease R-te: 1.20E+OOS uCilsecMlethod of Projec don:STAMPEDEOffsile Dose Projection (rem):ImileTEDE 0.072CDE 0.506-miles0.0250.175miles0.006OL04110miles0.0020,013Projected duration ofreleae: h.O thourA General Emergency Requires a Protective Action RecommendationEVACUATE ZONE(S): 1SHELTER IN PLACE ZONE(S): 2AFFECTED DOWNWIND SECTORS: R. A, BAll Remaining Zones Go Indoors And Monitor EAS Radio StationBased on a Site Boundary (1 Mile) Dose Projection > 0.1 rem TEDE and/or 0.5 rem Thyroid CDE theEmergency Classification Initiating ConditionRSl (SITE AREA EMERGENCY) has been met.PERFORMED BY:REVIIWED BY:RadiiiLd.tntger.,'.adiological DirectorI 17T2013 3:29:00 PMDaterime.Daterime12'17fl013 3-2&:53 PM CALC. NO. STPNOC013-CALC-002Radiological Release ThresholdsEN ER CO N for Emergency Action Levels REV. 2Excelence--Evey projec. Every doy. Attachment 3 PAGE NO. 42 of47DRILL STAMPEDE User Supplied InformationDRML PxW-47.0-33 9z2f01W1DRLDa mdt e : 12d17i2013 15:26C-nmisur fNam: Uni 'it (ekmwl=au SulitdmztILbmrtqwdeDatallhhopmt:Gr(mmudlk aMdwlo-ity 131 mit/hrQound IuL tdfranr 10 &O greeUer-e ttedStahly ClmSlaby clas: "D -NetufriItndtme~dUhl Vent Rinse:11oi Vent eRate terae 2.50E+CM nCi/sec1Reat Shuttun Dat/lutme: 12/1712113 14:26Rinse Start Datg/rTi: 12117/2013 15:26FltinmtedRolmsefDum-io- 1-00 hImrsNucdi Mhlnre:ClaphntnqyCauIcuedNOBLE GAS rne rate: LiB'O u -ilsecNOBLEGASNudkide m/sKIODMIENuduikb RUiSKPAR3fCL.AUfNucL& mCI/sacIr-tmKY-IS:1K,--45MKr M7:lKr-4S:K,,r-It.XTe-lW2M2E+0061.0-,M+0067.0Z2+0069.03E+_01.7313+0073.1(E+0013M22E+0X76=l+007191E+al61.45E+0X179.6fE-0061-131:1-132:1-133:VIM:1-L35:3.1914M3.0811+005.12E-405Cs-lid:Cs-137:celpr-lU:Ce-141:La-lID:11,1-99:Nm/Rh-l~dLSr/y-w:Sr-":Te-132:Zr-95:1.05E+aD26.25E-(X)I82.10022.1E--002531EOD3.43EWCC1.25FA-0B9.1CE-0031.321-0013.13M-0121q742013 3:2637 PM CALC. NO. STPNOCOI3-CALC-002~ E N E R ~ NRadiological Release Thresholds REV. 2C 0 N for Emergency Action LevelsY-pr. Attachment 3 PAGE NO. 43 of 47DRILL STAMPEDE Results InformationDRILL %Pagelaf 2 DRILLDaftznrm 12mfl2013 1526 UsNlk ne: CIEMMa,mflur I~aftgDixtnsa051.02.05-n7-510.020.0Distuca051.02.05.07510.?20.0Diunca(siles)0-51.0205.07.513.020.Plus. bad Min.(hecunsmiurths)01150-230-45131.C111V YangJL032E-0053.755-1"06LiME-GOW5.7DE-077I1-W1E-007CHUQ UM'2.43SE-09-11tH-GM31521E-0IM7372-0072-441E-009.109E-00&easunde DoseRtIsrieu flu! 5e4m0z79033220,117C.C290.0160.0100.00MImmr"ion Wfk.le B-dynabul gas gmma (rum)0.8790-3320.1170-M90.0160.0100-03MEm lodiza CD,exerrsl + itenlau "lyraid(remskr) (nsir)1.5_ 13.640.601 5090210 13,620.050 0.4110.027 02140.017 0.1350.006 OM05ME bidine C DEcuornu!l + jut.rual ¶Ikyri-d(,rum) (res)1.5980.60102100.0500.0270.0170.00613.6465(3M1-7620L41102140.1350.0501211..013 3 2 625 YM CALC. NO. STPNOCOI3-CALC-002Radiological Release Thresholds R EV__2J E NERC0 N for Emergency Action Levels REV. 2Fxcelence--Ever/project. Every day. Attachment 3 PAGE NO. 44 of47STA-MiPEDE Results InformationDRILL Rev-ion7Oit3 92Wi3011 Page 2of2 u DRILLC.cultaliorn CompltedRESULTSWindVelocity: 132ni/hrWindflirection: 120Release Rate: 1.19E+-0O uCisecAfethod of Projection:Offsite Doe Projection (rem):I mile 2 mile 5 miles 10 milesTEDE 0.601 0210 0.050 0.017CIDt 1-62 0.411 0.135Projected durta-on ofreleme: 1.0 hoursA General Emergency Requires a Protective Action RecommendationEVACUATE ZONE(S): 1, 2SHELTER IN PLACE ZONE(S); 6. 11AFFECTED DOWNWIND SECTORS: R, A, BAll Remaining Zones Go Indoors And Monitor EAS Radio StationBased on a Site Boundary (1 Mile) Dose Projection > 1 rem TEDE and,/or 5 rem Throid CDE theEmergency Classification Initiating Condition RG1 (GE NERAL EMERGENCY) has been metPERFORMIED BY:NameRfMIEW-ID BY:Kad .iinagerRadiological Director11q 7/2013 3:26:33 PFMDateflimeDaoe.Time121 7,2013 326:25 PM CALC. NO. STPNOC013-CALC-002r Radiological Release Thresholds REV. 2IAeO E for Emergency Action Levels 3EN.4F.xce.,flnce-E-Evpy profect. Every day. AttachmIlernt 3 PAGE N O. 45 of 47DRILL STAMPEDE User Supplied Informaion DRILLDRMDRILLm.-33 f&qIDhhilme: 12q7,2013 1530CUMMnRhe-r Namme: SteamLimeIf-rmo ndll ,tadwabc.ty 1-2 ni/krQacnlmkewdi dfrimn I0 yCIlassStibilty Chus: fl -Nmenr"3tnihorMdSJGTuhe Raipiur debasc:St 45-M+@02 uCiWcStem fl=wRate: LONO mllrOBactar Shut&= Datwflme: 12/17VIh13 14:30Rmnse Stat BatWklma: 12/17l2013 15:30BMtantedR1ease Duratiar 1. kournNmrh& 1skim:e NoM. Gas +Jmhtelotfm as pwc-rt ofnade gas: C26LNO.LECAS reeae rute: 12 009 nCi/secNudILEnC/SNudide IODMLENudide MiUMePARIICLAIENurfida aCilsecfl-ISMKr-IS:Kr4ffMKT497:KY-ItKr-StXe-l31n&xfL&.12-133k!XLe-IS:Xe-13SSXL-137-12u?-Mt1.14E+0063.45E-IXT12SM4U06925E+0064.13E-01127E4UJF73.LqE-041712M1+005124F-+00!135E'+005.-131:1-132:I--1]:1-L03:1-134:V1,.5:3.24E-054-QE40054214H+00tCsr-134:Cs-137:Ce)Pr-1dA:Ce-141:La-iS0:Mkb-99:MdMR-106:En-l~l:Sr/Y-gI:.Sr-ID:0.IX4OO.DEE+CX)o0.04E+C00OAXE4000O.D0E+000,O(E+OO00.UfE-NXO00.(XE+000O.4E+00Te-132: O.X+CCO0Zk---: 0.CCE+O012A1712013 3:3035 PM CALC. NO. STPNOC013-CALC-002Radiological Release ThresholdsF4ENERC 0N for Emergency Action Levels REV. 2Fxceflence--Everproject. Every day, Attachment 3 PAGE NO. 46 of47DRILL STAMPEDE Results InformationDRILL7.0dskm23&3 t2O11 Pge lo2 DRELLDN rnr 12172013 15:30 Us P&M: Stam.n CGEMMmts:(miles)051.02.05.0751031.02.05.07_5*(miles)0.51.02.05.0751.102CLOMnPlume Trarel S.m(iunrsrmihetes)0:020050:230-340:45131CULtI VYang(ser/m/)I.03MM003.75.-54-065.004-0063-ME-007154E-oCHUQ DI1PL2A43d-0059.11CE-,W3.151.E-.D7.373E-003.,845,-007i2441E-0079:109.E-I ra. Dose R Iioneniea lde Body7abl0 zs pummz(0DEm/k)1.1090L42360-1530.0400JO220.015Immersion Whlel Bodynoble jas p tmms (rne)1.109OL4240.1530.0400JO220.015-OM6LPAG Dos RatesMEh Iodinea CDEextersal+ intiruer. fyroi'(rem/kr) (remir)1.2950.71702540US0.03400220.008135375.1!581-7470.2110.1340.049MME ZIodine CDEextrnal + intnrnal Murvid(remn) (rem)18950.717D02540.0630.0340.0220.008135375.-1.74702110.1340.049 CALC. NO. STPNOC013-CALC-002Radiological Release Thresholds REV. 2EN E R C 0 N toraEmergeny ActionE Levels REV. 2Fxpene Eey pojet vefydoy Attachiiieiit 3 PAGE NO. 47 of47PEDE Results InformationDIRILL ,Rion 7.03.3 9-2&2011 Page2of2 DFRLLCalcultions CompletedRESULTSMedthodofProjection: WindVelocity: 132m.,ai Release Rate: 120E+009 uCi'secSTA-MPEDEWindDirection: 180Oite Dose Projection (rem):1 mile 2 miles Smiles 10 milesTEDE 0.717 0254 0.063 00o2CDE 1,747 0.407 0.134Projected duration ofrelease: 1. hoursA General Emergency Requires a Protective Action RecommendationEVACUATE ZONE(S): 1, 2SHELTER IN PLACE ZONE(S): 6, 11AFFECTED DOWNWINXD SECTORS: R., A, BAll Remaining Zones Go Indoors And Monitor EAS Radio StationBased on a Site Boundary (1 Mile) Dose Projection > 1 rem TEDE andfor 5 rem Thyroid CDE theEmergency Classification Initiating Condition RGI (GENERAL EMERGENCY) has been metPERFORMED BY:Rn=7D BI. :NameRNlEW1TD BYDBad ihanager.,Rmdlological Director12/1712013 3:30&.48 PMDate,/Tfie.DateMime1217/013 3:3039 NP CALC. NO. STPNOC0 I 3-CALC-004ID E N E R C O N CALCULATION COVER SHEET REV. 2.xcefl>n fe--ery ptoP G N Every" daPAGE NO. I of 17Title: Fission Product Barrier Failure for EAL Thresholds Client: South Texas ProjectProject: STPNOC013Item Cover Sheet Items1 Does this calculation contain any open assumptions that require confirmation? (If YES, Identifythe assumptions)2 Does this calculation serve as an "Alternate Calculation"? (If YES. Identify the design verifiedcalculation.) Design Verified Calculation No.3 Does this calculation Supersede -an existing Calculation? (If YES, identify the supersededcalculation.) 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 CalculationSafety-Related Non-Safety Related(Print Name and Sign)Originator: Caleb Trainor Date: 2/16/2015Design Reviewer: Chad Cramer Date: 2/17/2015Approver: Marvin Morris , Date: 2/17/2015 ENERCON CALC. NO. STPNOC013-CALC-004Eceienrp-Ev ey eery CALCULATIONREVISION STATUS SHEET REV. 2PAGE NO. 2 of 17CALCULATION REVISION STATUSREVISION DATE DESCRIPTION0 03/04/2014 Initial Issue1 03/21/2014 Updated the source term2 02/17/2015 Designate as Non-Safety RelatedPAGE REVISION STATUSPAGE NO. REVISION PAGE NO. REVISION1-8 2APPENDIX REVISION STATUSAPPENDIX NO. PAGE NO. REVISION NO. APPENDIX NO. PAGE NO. REVISION NO.A 9-17 I CALC. NO. STPNOC013-CALC-004E N E R C O N Fission Product BarrierFailures for EAL Thresholds REV. 2Excelen---vry project. Every oy F PAGE NO. 3 of 171. PURPOSE AND SCOPEThe purpose of this calculation is to determine whether a breach of Fission Product Barriers hasoccurred based on containment radiation levels. The calculated levels will be used as threshold valuesin 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 beanalyzed 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 CONCLUSIONSThe results of this calculation are listed below.Table 2.1: RT-8050 & RT-8051 Response toFission Product Barrier FailureFailure Monitor Reading (R/hr)Reactor Coolant SystemFuel Clad 40.2Containment 383Readings on these monitors at or above the values listed indicate a failure of the associated fissionproduct barrier.3. REFERENCES3.1. 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 toLOCA and FHA for South Texas Project Electric Generating Station, Rev. 3.3.3. A41009-00458UB, Unit 1 RCS Volume and Temperature Assumptions for Evaluation ofRadiation Sources, Revision 63.4. Offsite Dose Calculation Manual, Rev. 17, March 2011.3.5. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design BasisAccidents at Nuclear Power Reactors, Revision 2.3.6. STPEGS Drawing 9C129A81105, Radiation Zones Reactor Containment Building Plan at El. 68'0", Revision 33.7. MicroShield 6.203.8. 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. 1583.12. NOC-AE-07002127, Request for License Amendment Related to Application of the AlternateSource Term Regulatory Guide 1.183 Standard Review Plan 15.0.1 CALC. NO. STPNOC013-CALC-00410 E N E R C0 N Fission Product BarrierFailures for EAL Thresholds REV.2Excellric-Fvery projc Fvry doy, F r oe l PAGE NO. 4 of 174. ASSUMPTIONS4.1. Monitor LocationThe Containment High-range Area Radiation Monitors (RE-8050 and RE-8051) are locatedinside the Reactor Containment Building (RCB) close to the outer structural wall. The monitorsare mounted approximately 5 feet above the 68'-0" elevation (Reference 3.6). Limitations of themodeling software used (Reference 3.7) requires that the geometry be configured with themonitors placed immediately outside the source volume rather than inside the source volume.This slight difference is assumed to be negligible.4.2. DispersionThe entire reactor coolant system (RCS) mass of noble gases are released into containment andevenly 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 inReference 3.1 developer notes.4.3. BuildupBuildup was ignored, as the detector is immersed in the atmosphere. This is conservative, as itproduces a lower dose, which provides a lower detector setpoint level.4.4. ActivityThe activity of the RCS corresponding to the failures of the RCS, fuel clad, and containmentbarriers are Technical Specification limit, 300 ýiCi/g DEI, and 20% fuel failure respectively asspecified 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 whichresults in a conservative estimation of 1200 [tCi/g DEI for 20% failed fuel.
OENERCONExcelence--very projct. F~vey Fission Product BarrierFailures for EAL ThresholdsCALC. NO. STPNOCO13-CALC-004REV. 2PAGE NO. 5 of 175. DESIGN INPUTSThe normal operating coolant inventory (for noble gases and iodine) is taken from Reference 3.12Table 4.2-20 and is listed in Table 5.1.Table 5.1: Reactor Coolant InventoryNuclide Activity (iCi/g)Kr831m 3.70E-01Kr85m 1.50E+00Kr85 7.60E+00Kr87 9.80E-01Kr88 2.80E+00Kr89 8.40E-02Xe131m 2.80E+00Xe133m 4.20E+00Xe133 2.40E+02Xe135m 4.OOE-0 IXe135 7.60E3+00Xe137 1.60E-01Xe138 5.80E-0 11131 4.25E+011132 6..00E+01133 7.OOE+011134 1.30E+011135 1.90E+02The dose conversion coefficients for iodine are taken from Reference 3.8, Table 2.2 and are listed inTable 5.2.Table 5.2: Dose Conversion FactorsDose Conversion Factors(Sv/Bq)1-131 8.89E-091-132 1.03E- 101-133 1.58F-091-134 3.55E-111-135 3.32 E- 10The containment free volume is taken from Reference 3.2 item 3.2 and is 3.8x106 ft3.The mass of the RCS is 2.658xl 08g per Reference 3.3.6. METHODOLOGYEquation 6.1 shows the calculation used to find the release concentration of each nuclide. CALC. NO. STPNOCO03-CALC-004l E N E R C O N Fission Product BarrierFailures for EAL Thresholds REV. 2 of17Exce en e-F ery pm e Cve Ey doy PAGE NO. 6of1(XI-13,*DCFI-131]Euto .Release Activity =Ci
- MRCS
- Equation 6.1Si=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 specifiedactivity,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), andIi 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 todetermine release activities. The STP ODCM dictates that the DCF which define DEI are to be takenfrom Reference 3.8. Using these conversion factors, the dose related to I ýtCi/g, 300 pCi/g, and 1200pCi/g is calculated.The dose rate from 1-131 is then compared to the dose rate from the full iodine spectrum in the reactorcoolant inventory found in Table 5.1. A ratio between the two dose rates is used to scale the coolantinventory 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 equivalentvolume to the containment free volume. This is filled with the evenly distributed source activity with adetector 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 byENERCON Services, Inc., MicroShield 6.20 Computer Code Verification, STPNOCO13.7. CalculationsThe spreadsheet below contains the calculations described in the methodology section.
CALC. NO. STPNOC013-CALC-004O ENERCONExceflenc--Everynprojec. Eveiy doy,Fission Product BarrierFailures for EAL ThresholdsPAGE NO. 7 of 17Table 7.1: Release Inventory for Failed Fission Pr(NuclideKr83mKr85mKr85Kr87Kr88Kr89Xel31mXe133mXe133Xe135mXe135Xe137Xe13811311132113311341135ConcentrationpCi/gm3.70E-0 I1.50E+007.60E+009.80E-0 12.80E+008.40E-022.80E+004.20E+002.40E+024.OOE-017.60E+001.60E-015.80E-014.25E+016.OOE+017.OOE+011.30E+011.90E+02Gross ActivityptCi (concentration*mass)9.83E+073.99E+082.02E+092.60E+087.44E+082.23E+077.44E+081.12E+096.38E+101.06E+082.02E+094.25E+071.54E+081.13E+101.59E+101.86E+ 103.46E+095.05E+10Ci @ 1 I Ci/g DEI-131Ci total=activity* Scalingfactor/1061.57E+006.35E+003.22E+0 I4.15E+001. 19E+OI3.56E-O11.19E+011.78E+011.02E+031.69E+003.22E+016.77E-0 12.46E+001.80E+002.54E+002.96E+005.50E-018.04E+00oduct BarriersCi @ 300pCi/g DEI-131Ci total=activity*Scalingfactor/1064.70E+021.91 E+039.65E+031.24E+033.56E+031.07E+023.56E+035.33E+03_3.05E+055.08E+029.65E+032.03E+027.37E+025.40E+027.62E+028.89E+021.65E+022.41E+03Ci @ 120OOACi/g DEI-131Ci total=activity* Scalingfactor/1061.88E+037.62E+033.86E+044.98E+031.42E+044.27E+021.42E+042.13E+041.22E+062.03E+033 .86E+048.13E+022.95E+032.16E+043.05E+043.56E+046.60E+039.65E+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 0Excelence-Every project Every doyY.CALC. NO. STPNOC013-CALC-004Fission Product BarrierFailures for EAL ThresholdsPAGE NO. 8 of 17Nuclide11311132113311341135Table 7.2:................................bI e............A ActivitygCi/g4.25E+0 I6.00E+0 17.00E+0 11 .30E+011.90E+02Full Iodine SpectrumDCFSv/Bq8.89E-091.03E-101.58E-093.55E-113.32E-10Total =...... ~ ~ ~ ~ ~ ~ ~ ~ ~ ... .......................tv t
- CActivit y DC F(pCi/g)*(Sv/Bq)3.78E-076.18E-091.11 E-074.62E-106.3 1E-085.58E-07Table 7.3: Scaling FactorsDose Equivalent Activity(ýtCi/g DEI)3001200DCFDEA*DCF(Sv/Bq)8.89E-098.89E-098.89E-09(paCi/g)*(Sv/Bq)8.89E-092.67E-061.07E-05Scaling Factor(DEA*DCF) /(sum(Activity*DCF))1.59E-024.78E+001.91E+01 Fission Product Barrier CALC. NO. STPNOC013-CALCO004r EN ER CO0N Failures for EAL Thresholds REV. 2Exelene-I'vy polct Fvefyda Appendix A PAGE NO. 9of 17Case Siumainy of Fuel Clad Barier Page 1 of 3Micro!'hield 6.20 (05~-MSD-6.2O-1158)EnereronFe~M 1Re2f1DO Fil I,-,.Cftecked_Case Tile-- Fwel Ciad BafTi~rDescriptiont: ! P0CGzCGeometrin 8 -CyI djr vIulufi -End1 SIIeI1s14*1gMt ~ .e3 m (192Ft 11 5 n"Radius .&3c (7E ft)Dot-e P&ýntaA X YI 50ý.6 CM a 112,64 cmý I j c2 0it ;,,I 1i93 110 0 in1 j _7ftE _- __ýSN~ew N DimesloaI HAMatai Deni*tyS,"r; 156-s10 C1, Al .012Source Thput 91 Gu ngj S-a '_w: IcN of G~p 21ULAeer *Ile1'gv Ciulo : .015phoýtmw,3 < 0.015 I lnrudadT,132 2.4Q+09 0 .01 2 36-09732014 1-1233 2.65+Q 1.92 11 3 0t55e-X05 .33÷1_13A 5.5000me- 131 2.035 Ue + D1, 55ý,629 07, 21 795-00G1-135~~ CG e + 0400+0 2,9748a+0i *D I9s0 I6I 65n 00ýe+CZ00 2.349,75E +011 6.5765e-N_0S Aj4323e+0IKr-87 4.1 b(+ 1_!53 55a7+ i11 4 298G0e-005S 3C 359e3+ ,v,11 t-81. 17 2,?,+ 10 3. 6r,7 C -0C6 1.35642.2-0,71Xe-lImIIa 1. 1 KC la+ 0 4.45+ 11 .34-0 4.S9+0)(*3 3 I 1 2, I 3774?11LJ13 I 05944e-CX02 3.90564I+X002Xe- 333m 1.7890af_4+C01 65.5560f5+011 I .6435e-OC4 b8209e+1000Xs- 135 3,22F004+C'0 1.19 Sa+3!2 3 33111-004 .2339a4+001Xý- 13 5m ,P I.9 0e +CX0 62530s+010IC .7503es-005 F 6,76f*-o00Xe-137 6.77D0s-001 7.50C9eI01I7.0-15X-f, 2 .5942 e- 00u1X I I t2.4C -00+ý" ,0 Y. 1 02e4-" 1 2 54 7 *-Q X 9;42566-001BuldpTh atrllrdeeic s oucEbI- ~ N~ Fhaenom Rate Fluevee Rate Exposiare RAte Exposure RateHeY htot/ mevpvn3/c aC Pmev/cm.laec mR/hr mth#4 Eufu With Uufldap No sul~duP With 11Mu~dp40.015 2.437ie+12 4 1. 29 1!z F2 1 A 53e , (2 .1 07,E+O I 2461+01file -MCJProgran%2OFiles%/2 O(x86)/M cro ShieldlExamples%/Casefil es1thflA%1/20 gC L. 2117/2015 Fission Product Barrier CALC. NO. STPNOCO03-CALC-0040 E N E R C O N Failures for EAL Thresholds REV. 2Fxcellence-EFveyprojet EveFy yd'y Appendix A PAGE NO. 10 of 17Case Summary of Fuel Clad BarrierPage 2 of 30-430.60.11-00.520.30.4ATOWS1.880e+132.13 le + 11. 804e +112-370e+112.069e + I2.644e + 11J , SOe+1I2.605e ,105. 89B + 033.841e+23I.03?L+003.852~e-023.232e,033. 144 &-02I. G4 3.e-031. 419t -03L. 459e-032. 604e+032. 930e- 035.755e+0U31. 0 1 E+041. 1496+3&-C4.860e+04S. 1 45e032,3130e+041, 70 5- 005,761e+024.394e +033 .9686+021 261e+031 .66'e+(131.65r0e+032.911e+033.17.4e-036.156+e031.0716+041.198e+036.55'3e+046-.344e-015.964e-01,2-032&+r0O2-785e+C-D2,S817e6,0O4-9526+C05-309'e+001.sm.e+c014.334".28.072e+013.640ei-017.755e+007.52 7e-012.45ie -DG3.271e+005.548e+0,35.s551+001.041e+011 .857e+011.626e+004.483e.021.8816+02f~ie:///CJlProgram%2OFiles%20(x86)fMcaoShield'ExamplesICaseFiles/HTh1/1%20-)iCi.. 2/17/2015 Fission Product Barrier CALC. NO. STPNOC013-CALC-0040 E N E R C O N Failures for EAL Thresholds REV. 2Exc,,nre-E'project, ,vetydoy Appendix A PAGE NO. 11 of 17Case Summary of Fuel Clad BarrierPage 3 of 3to.,zfile..f//C:IProgramlc2OFiles%2o-O(x86)/MfiaoShieldf-xamples/CaseFiles/HTMEl/%2OgiCL.- 2/17,'201 5 Fission Product Barrier CALC. NO. STPNOCO13-CALC-0040 E ENE R C O N Failures for EAL Thresholds REV. 2nellec-Fry-roj- Evef day. Appendix A PAGE NO. 12 of 17Case Sunmmary of RCS BarrierMicroShield 6.20 (05-MSD-6.20-1158)EnerconPage 1 of 3pa'geDOS FileRun DateRuan TlmeDuration-t300 WC, DEI.mti6~Fcbruary 17, 012ýS3A45 PMM 0,:0:Do0File RerDatatByChe-ked(As TfflIei R.Cs BarrierDescriptiom 300 PCI/cc8 ffeo -Cy!i~ra8r Vdolume -End SýlIeldsSI~urce DlnterltonE$Height 5.9c+3 cmRadius 2.3e+3cmf 192 ft 11 .5 in)ý7sfrt)A*1x y z601.6 CmI 582.64 cm 1143 cm20 lt 4)[)in 11;3 FT00 In 37 It 60 ýShieldsShield N Ditmensimi materilasouice 9.66+10 CM3 AirAil Gap AXTSource Input -Groupinlg Kettbod -Standard LatticesNumber of GPMups 25Lower Ewgy Cutoff, 0.015pwunts < 0.015 " adaldedUllrary : rv0.001220.00122NucitdeI- 13t1-1321-1331-134Xe-137Xe- 1335.40001e-a-0247.600r8+0021.65008+003ý1. 91008+0 D 31,:40rt!-M33.508+01,07008a+0023. koc0e+QC`3.05008+e'005UC30e039 f-S008+0035 r08008+022 0300e+ý0027 370Ce-002buecuerel1.9980e+(0132,8194e+013ý332;938+0137 -0-57Ce+ G1335705se+016L1 72le+C1417261)2e,313-S.5926e-003i7,8918e-0039.2071e-c()]1 7088e-C032.496r08-0024 8676e-C0039'9942e-0021978 18-Q023 687rje-0023.6870e-0023 .1588 e+ý0005.5201e-00299942e-GO2S2612ii-0032.10248 0037 532;98-0032.0693e.0022.9200800923.4ý0668002c.3227e.+0010037,3-'19 18+ DO24 7518ý,+ 021-3612t+01)34,1002e+101.1.M42e8-0031.1687t-+1052,04248+003ý3.69788+00C31.9486"80027.77898+ý00112.8242e.002BLIldup : The maaittLV refaelele IS -SoUfceIneoM" ,rameteis20RadialCOecumnlerfIlliaI1t10Enetgy04eV0. .......015ACtivYty Flpulca RatePhorara/ac l~eV/O~n2/**7.436e-14 ~AI &* 8+9 4FteritsFluence RateNev/cmli/secWith Buildup4.34le-04Exposure RatemR/hrNo Buildup3.311e+03Exposure RatemR/hrWith Buildup3.726e.+03ffile:W/C /rogram%,2OFilesl20(x86)!NlicroShieldVExampkesiCaseFiles/HTMLJ3O00%2OpiC. 2117"2015 Fission Product Barrier CALC. NO. STPNOC013-CALC-0040 E N E R C O N Failures for EAL Thresholds REV. 2F vey da Appendix A PAGE NO. 13 of 17Case Summmry of RCS BamerPage 2 of 30.030.080-.15020L3o4.01.55.622e+ IS4. 142e -152.598e -116 .3 10e +133.8 7Be+ 142.448e. 135.954L 136. 399 e, 1-35.409e 137. 104e +136.200e 138.070e 131.ý047e+147.787e- 12-1,773e ,111.149e+16I .539e +063.ft96+063.0O886+,-G1. 158e,+C59.683e+059.437e+043. 123p, 0,4.2F 16+054. 375f,-,057,803e,+G58,6314+05ý_1. 724e- (,,3.027e+063. 434e4 0-51.053e+04IA55e+072,436,E-061.7 1 e+0D53 774e+055.004R4055.033-+058 742e,059.514e+051. S54,4ý0,53.206e -OF3.532-*051.689, +041.960e+07I 525.e+046.123e +G31-.907e+021.7G96+03I 790e+026.0854a+029 363e+028_54Ge+021.484e+031 .591e++32.901e+034,681e+034.65.9e+021_303e+014.020e+042,414e .04I .088e+047.8G3e-012.85 1e+022,323e+032.259,R +027,354e-029 B22e+029.834e+021,.663e +031.754,-03TI 196+034.958e+0134.660e+ 0'1,347-+015.68e8+04-file-./!C-/Pogr'am0/*2OFiles*/20(x86)MciroSbiield/ExamplesICase~iles/HdflJ 3O0%*20gC... 211712015 I Fission Product Barrier CALC. NO. STPNOC013-CALC-004E N E R C O N Failures for EAL Thresholds REV. 2E Everyday. Appendix A PAGE NO. 14 of 17Case Summary of RCS BarrierPage 3 of 3zffiecI//CilPrograml2OFilesl2O(xS6AicroShield/E-xamplesiCaseFile~fl1Tfl30OO%2OjiC. 2/1Ti201 5 Fission Product Barrier CALC. NO. STPNOC013-CALC-004E N E R C O N Failures for EAL Thresholds REV. 2Ecelnce-FEvrypret Evedy Appendix A PAGE NO. 15 of 17Case Summary of Contiamnent BarrierPage 1 of 3MicroShield 6.20 (05-MSD-6.20-1158)EnerconPageDOSFileRun Date,Run TimeDuration~1200 WCI DEL.ms6February 17, 20152,41 11 PMý.0 -0~0 0FI'le RA"CheckedCA"O Tltleý Coannment R5~fierDegcrlptjoe 1200 pci/cc6oojetiy. 8 CylInder Volumne -End hedsouroi, Dimenst~nsýRadius 2 3e+3 crrn J'7ft)voý,e PointsA X Y61019.6 cmi 5SI3264 cm, 1143 cm20 ft 0- In 193 TtO0.0In 37 R 6.0iWShield Nsou ceAC;ApslhfeldtDimension~9 66ce+ 10 cfrS"I'M Input. GroUF~ng Method -Standard radicesMumber of Groups 25Lower Eney Cu~toff~ 0.01sPhotýn S < 0.0 15 -Ind itJedLib~w,, ý GroveNucilde1-1311-1321-1331-1341-135Kr F83miKr- 3Kr-S 7Kr4S8Kr 89Xe-131wXe-133Xe-133muXe-135Xe-135mXe-1372. 1600e413043.rO5LK'e+0C,13.58E. +004-rQ,616000e+0031388(0C*0037.62OCý" 003lAaK0e+ 0G44.2700e+0021,4200,e-0041 .2200e. Out2.0130ce -+008. 1 300e+0022.950r~ep007.9923e+0141, 28q5e+0152.442DC-0143 'i705,4.0156.95(H05+013I 4282te+GIS2,519'4e.s014!.842t-e+C145S40+1L57995+,1135.25405+0144_5140ýe+015788310etf3!41.4282e+L157.51 1t0+0133.D0051a0131,0915e+-0143.2370ie-V016.8354e-01129,99421e-001.94 71e 0023,9977,ý-0017.869 1Se -CC2S. 1576.e-cl021 .47,_,e-0)14 4223.e-0031 .4706-ke-W2.20505t-003 .99'7e.-C'M2. 1024a-r0028.4.,2005e,-V,33.0552e-Cr02Material DensityAir 0.100 1221. 1679li-O02.520015+0031,9083e- 0035.44114e5+0034.F750.54.055.1621.5-031.4791e-QC47.77859e+0023. 11541w,0021. 13 04 C -00ý3Buildup ;The matefia reference Is -Sourcelnl,9 raitio~n Parameters,Radlat 20YCIrcutetisI 1Enearg yMev.....ACMKVit Flue"Ce Ratephafne/ve "v/CrI/eN2o Run-up29 1e+ 1ý S47e.+i05Flaence RtateMeNF/Cm./eerwnh 74 le+0Exposuere RateMR/fhrno But*1d9p1. 327e+-AExpojure Ratema/lwWith Buldup1 443e+04fiaei//lCirograml,2OFIks%2,?O(x8)MicroShielVExamples/CaseFiks/1TM1J1200%2OL..ý 1217/2015 Fission Product Barrier CALC. NO. STPNOC013-CALC-004SE N~ E R N Failures for EAL Thresholds REV. 2Excellence-Every prqect. Evy day. Appendix A PAGE NO. 16 of 17Case Summary of Contianment Barrier Page 2 of 30.03 2.254e16 6.172e,- 06 '9,767e-06 O.117e+04 9.679e-040.6 1 -6f$9e 16 ..1.5414e 075f+4 .5e00.1 1,03ý6q12 1.232--03 2. 34e÷03 1 4+ 00 3. 1 2e4000.15 2.647e -14 5 7.265+÷0D 8.00,e+02 1.196e+030...2 1.624e- 15 4.055, 06 ...13. .06 7-157e+03 9.730e+030.3 2.1)33eý14 1.131e-06 1.427e+0,6 2-145e+03 2.707e-03D04 1 0," e ,15 5 .4 7 9e"G6 6 b214+065 1.067e+04 I 24ce+040.5 1.20e+15 I 212e+07 1.423e-07 2 794-e040-6 1 ý679te+,15 1. 358e, 07 i34+7 2-651e+04 3.1)53e- D40.C8 1. 98 5e -I 2, 11e+07 2,443e +D7 4.14144-04 4.1,46e +041.0 2.769ea-07 3.053+07 5 105e+04 5.627e.0415 2.2768e+15 4.8684+07 5 235e+07 8.1544+04 .0e42.0 8.228e+14 2.377e+07 2.518e÷07 3.89et-04310 3.124e+13 1 .377e+06 1. 437e+06 1.869e+03 1 950e+034.0 7..75......... e+1 ..2e...4 4,34F..04 5 19a,-+01 5.3 +o01Totals S.Sg84+16 1,820e+08 2.156'e+08 3.531*4.05 4.721e+05file:/IIC~Ilrograml2OFilesl2Ox86)/MicroShiieldExamplesICaseFiles/HTML/1200%2Oiý.. 2/17/2015 Fission Product Barrier CALC. NO. STPNOC013-CALC-004E N E R C O N Failures for EAL Thresholds REV. 2,yroiect veryday Appendix A PAGE NO. 17 of 17Case Summary of Contianment BarrierPage 3 of 3K.file-./I/C:Program%/ý2OFiles*/20(x86),McroShield/Eamples!CaseFilesIHIMJ 1200%/204... 2117,2015
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