Text

Monticello Nuclear Generating Plant - License Amendment Request to Revise Emergency Action Level Scheme: Attach. 1, EAL Comparison Matrix Document (Deviations & Differences) and Attachments 2 & 3, Red-Line & Clean Versions of EAL Technical
	ML17095A112
Person / Time
Site:	Monticello
Issue date:	03/31/2017
From:	Gardner P A; Northern States Power Company, Minnesota, Xcel Energy
To:	; Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
Shared Package
ML17095A107	List: L-MT-17-012, License Amendment Request to Revise Emergency Action Level Scheme: Emergency Action Level Matrix, Hot Conditions (Attachment 5); ML17095A112; ML17095A113; ML17095A114; ML17095A115; ML17095A118; ML17095A120; ML17095A121;
References
	L-MT-17-012
	Download: ML17095A112 (394)
	v • d • e

2807 West County Road 75 Monticello, MN 55362 800.895.4999 xcelenergy.com March 31, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket Number 50-263 fl Xcel Energy* RE S P 0 N S I B LE BY NAT U R E L-MT-17-012 10 CFR 50, Appendix E 10 CFR 50.90 Renewed Facility Operating License No. DPR-22 Independent Spent Fuel Storage Installation Docket Number 72-0058 License Amendment Request to Revise the Emergency Action Level Scheme

References:

1) Nuclear Energy Institute (NEI) 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ADAMS Accession No. ML 12326A805)

2) Letter from Mark Thaggard (US NRG) to Susan Perkins-Grew (Nuclear Energy Institute), "U.S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 6, Dated November 2012 (TAC NO. D92368)," dated March 28, 2013 (ADAMS Accession No. ML 12346A463)

In accordance with the provisions of 1 O CFR 50, Appendix E,Section IV.B, and pursuant to 10 CFR 50.90, Northern States Power Company, a Minnesota corporation (NSPM), doing business as Xcel Energy, hereby proposes to revise the Emergency Action Level (EAL) scheme used at the Monticello Nuclear Generating Plant (MNGP) . . The proposed change will revise the current EAL scheme used at MNGP to the EAL scheme contained in NEI 99-01, Revision 6, "Development of Emergency Action Levels," (Reference 1 ), which was endorsed by the NRG in a letter dated March 28, 2013 (Reference 2). The current MNGP EAL scheme is based on NEI 99-01, Revision 4, "Methodology for Development of Emergency Action Levels," for security EALs and Revision 5 of same document for security-related EALs. The MNGP Emergency Plan, as changed, would continue to meet the standards in 10 CFR 50.47(b) and the requirements in Appendix E to 10 CFR 50. Emergency Action Level , f c . )JN 552 (o

/JJJJ s 5 Document Control Desk Page 2 scheme changes are required to be submitted for prior NRG approval as an amendment to the operating license as stipulated in 10 CFR 50, Appendix E, Section IV.B.2. The Enclosure to this letter provides NSPM's evaluation of the proposed change. Attachment 1 to the Enclosure provides a description of the deviations and differences of the MNGP proposed EAL scheme from the NRG endorsed NEI 99-01, Revision 6, EAL scheme. Attachment 2 provides a red-line/strike-out markup version of the NEI 99-01, Revision 6, EAL Technical Basis Document for MNGP for information only. Attachment 3 provides a clean copy version of the MNGP EAL Technical Basis Document.

Attachment 4 contains copies of the supporting calculations for EAL threshold values. Attachment 5 contains the Emergency Action Level Matrix, which is submitted for information only. NSPM evaluated the proposed change in accordance with 10 CFR 50.92 and concluded the change involves no significant hazard consideration.

Additionally, NSPM has determined the proposed change does not authorize a significant change in the types or total amounts of effluent release or result in any significant increase in individual or cumulative occupational radiation exposure.

Therefore, the proposed amendment meets the categorical exclusion requirements of 1 O CFR 51.22(c)(9) and an environmental impact assessment need not be prepared.

In accordance with 1 O CFR 50.91 (b)(1 ), a copy of this application, with enclosure and attachments, is being provided to the designated Minnesota Official.

NSPM interacts periodically with the state, county and local emergency management agencies.

NSPM has communicated with these agencies the intent to change the EAL scheme to the endorsed NEI 99-01, Revision 6, EAL scheme. These agencies have agreed with the change. The new classification scheme will be reviewed with the agencies once the license amendment request is approved and prior to implementation.

NSPM requests approval of the proposed amendment by April 02, 2018. Once approved, the amendment will be implemented within 180 days. Please contact Lynne Gunderson, Licensing Engineer, at 651-267-7421, if additional information or clarification is required.

Document Control Desk Page 3 Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

I declare under penalty of perjury that the foregoing is true and correct. Executed on March 31, 2017. Peter A. Gardner Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company -Minnesota Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Monticello Nuclear Generating Plant, USNRC Resident Inspector, Monticello Nuclear Generating Plant, USNRC State of Minnesota L-MT-17-012 Enclosure ENCLOSURE Evaluation of the Proposed Change License Amendment Request to Revise the Emergency Action Level (EAL) Scheme 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Proposed Changes 2.2 Background 3.0 TECHNICAL EVALUATION 3.1 NEI 99-01, Revision 6, Evaluation 3.2 Incorporation of EP FAQs 3.3 NRG Information Notice Evaluation NSPM 3.4 Radiation Monitor Differences Between Current and Proposed EALs 3.5 EAL Scheme Change Conclusions 4.0 REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination 4.4 Conclusions 5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

ATTACHMENTS:

1. Emergency Action Level Comparison Matrix Document (Deviations and Differences)

2. Emergency Action Level Technical Bases Document (Red-Line Version) (For Information Only) 3. Emergency Action Level Technical Bases Document (Clean Copy Version) 4. Supporting Calculations for Emergency Action Level Thresholds

5. Emergency Action Level Matrix (For Information Only) Page 1 of 13 L-MT-17-012 Enclosure 1.0

SUMMARY

DESCRIPTION NSPM In accordance with the provisions of 10 CFR 50, Appendix E, Section IV.B, and pursuant to 10 CFR 50.90, Northern States Power Company, a Minnesota corporation (NSPM), doing business as Xcel Energy, hereby proposes to revise the Emergency Action Level (EAL) scheme used at the Monticello Nuclear Generating Plant (MNGP). The proposed change will revise the current EAL scheme used at MNGP to the scheme contained in NEI 99-01, Revision 6, "Development of Emergency Action Levels," (Reference 1 ), which was endorsed by the NRG in a letter dated March 28, 2013 (Reference 2). The current MNGP EAL scheme is based on NEI 99-01, Revision 4, "Methodology for Development of Emergency Action Levels," for non-security EALs and Revision 5 of same document for security-related EALs. The proposed changes to the EAL scheme contained in this submittal do not reduce the capability of the MNGP Emergency Plan to meet the applicable emergency planning requirements established in 10 CFR 50.47, "Emergency Plans," and 1 O CFR 50, Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities." Emergency Action Level scheme changes are required to be submitted for prior NRG approval as an amendment to the operating license as stipulated in 10 CFR 50, Appendix E, Section IV.B.2. 2.0 DETAILED DESCRIPTION 2.1 Proposed Changes The proposed change will revise the current EAL scheme used at MNGP to the endorsed EAL scheme contained in NEI 99-01, Revision 6, "Development of Emergency Action Levels," (Reference 2). The current MNGP EAL scheme is based on NEI 99-01, Revision 4, "Methodology for Development of Emergency Action Levels," for non-security EALs and Revision 5 of same document for security-related EALs. Detailed descriptions of the proposed changes and supporting information are provided in the following:

Attachment 1 provides the EAL Comparison Matrix Document (Deviations and Differences), which compares the NEI 99-01, Revision 6, EAL scheme to the proposed MNGP EAL scheme. This attachment also addresses any differences between th8' endorsed guidance and the site-specific scheme.

Attachment 2 provides a red-line version of the proposed MNGP EAL Technical Bases Document.

This attachment identifies by mark-up the changes to the NEI 99-01, Revision 6, EAL scheme as incorporated into the site-specific EAL Technical Bases Document.

This attachment is provided for information purposes only.

Attachment 3 provides the proposed MNGP EAL Technical Bases Document.

Page 2 of 13 L-MT-17-012 Enclosure NSPM

Attachment 4 provides five supporting calculations for several EAL thresholds.

Attachment 5 provides the proposed Emergency Action Level Matrix, called a wallboard in NEI 99-01, Revision 6, that incorporates the new EAL scheme. The Emergency Action Level Matrix is provided for information purposes only. The proposed change summarized above is acceptable because the MNGP Emergency Plan, as revised, will continue to meet the requirements of 10 CFR 50, Appendix E and the planning standards of 10 CFR 50.47(b).

2.2 Background As noted above, the current EAL scheme used at MNGP is based on NEI 99-01, Revision 4, "Methodology for Development of Emergency Action Levels," for security EALs and Revision 5 of same document for security related EALs. The current MNGP EAL scheme was approved by the NRG in January 2006 (Reference 3). As noted in the January 2006 Safety Evaluation, Nuclear Management Company (NMC), the license holder prior to NSPM, had withdrawn a commitment to provide a supplement to reflect updated security EALs for MNGP. Instead, NMC was to evaluate revised security EALs based on the guidance provided in NRG Bulletin 2005-02, "Emergency Preparedness and Response Actions for Security-Based Events," and implement any changes to the security EALs under the provisions of 10 CFR 50.54(q), which was acceptable under the Reference 5 guidance.

The NRG Bulletin 2005-02 (Reference

5) security EALs were implemented with MNGP (NEI 99-01, Revision 4) EALs in January of 2006. The NRG Bulletin 2005-02 security EALs also became the template for the security EALs identified in NEI 99-01, Revision 5. NEI performed a comparison between the NRG Bulletin 2005-02 and NEI 99-01, Revision 5 EALs and concluded that the intent of the NRG Bulletin 2005-02 EALs were fully subsumed by the NEI 99-01, Revision 5, EALs. NSPM adopted the NEI 99-01, Revision 5, security EALs in early 2010. In November 2012, NEI published NEI 99-01, Revision 6 (Reference 1 ). The NRG formally endorsed NEI 99-01, Revision 6, as documented in a letter to NEI dated March 28, 2013 (Reference 2). NEI 99-01, Revision 6, addresses changes recommended by the NRG, along with enhancements identified by the industry.

NEI 99-01, Revision 6, represents the most recently NRG endorsed EAL methodology.

3.0 TECHNICAL EVALUATION 3.1 NEI 99-01, Revision 6, Evaluation An EAL comparison matrix (Attachment

1) provides a tabular format of the initiating conditions (ICs) and EAL threshold values in NEI 99-01, Revision 6 (left-hand column), along with the MNGP proposed EALs (right-hand column). The matrix also compares Page 3 of 13 L-MT-17-012 Enclosure NSPM the proposed EALs in terms of differences and deviations from the NRG-endorsed guidance provided in NEI 99-01, Revision 6. Items were determined to be differences or deviations based on the definitions provided in Regulatory Information Summary (RIS) 2003-18, "Use of NEI 99-01, Methodology for Development of Emergency Action Levels," (Reference

8) and supporting supplements (References 9 and 10). The RIS information defines an EAL difference and deviation as follows:

A difference is an EAL change where the basis scheme guidance differs in wording but agrees in meaning and intent, such that classification of an event would be the same, whether using the basis scheme guidance or the site-specific proposed EAL. Examples of differences include the use of site-specific terminology or administrative reformatting of site-specific EALs.

A deviation is an EAL change where the basis scheme guidance differs in wording and is altered in meaning or intent, such that classification of the event could be different between the basis scheme guidance and the site-specific proposed EAL. Examples of deviations include the use of altered mode applicability, altering key words or time limits, or changing words of physical reference (protected area, safety-related equipment, etc.). NSPM evaluated and compared each individual EAL to the guidance in NEI 99-01, Revision 6, and documented the comparisons in Attachment 1 , the EAL Comparison Matrix Document (Deviations and Differences).

The only deviations taken were those deviations approved through Emergency Preparedness (EP) Frequently Asked Questions (FAQs). Emergency Preparedness FAQs incorporated into the MNGP EAL documents are discussed below and in the General Notes section of Attachment

1. The proposed MNGP EALs are contained in Attachment 2, Emergency Action Level Technical Bases Document (Red-Line Version).

This document is a markup copy of the NEI 99-01, Revision 6, EAL scheme with the MNGP plant-specific indications, parameters and information incorporated.

The proposed MNGP EALs are also provided in Attachment 3, Emergency Action Level Technical Bases Document (Clean Copy Version).

Attachment 3 is a clean copy of Attachment 2 and will be incorporated into the .site's MNGP EAL Technical Bases Document following NRG approval of this amendment request. 3.2 Incorporation of EP FAQs Where appropriate, information from EP FAQs has been incorporated into the* Attachment 2 and 3 MNGP EAL Technical Bases Document.

The EP FAQs are noted in the Generic Differences and the General Notes of the Attachment 1, EAL Comparison Matrix Document (Deviations and Differences), and in the specific EAL justifications.

The specific EP FAQs incorporated are: Page 4 of 13 L-MT-17-012 Enclosure NSPM

EP FAQ 2015-003, regarding Revision 3 of Boiling Water Reactor Owners Group (BWROG) Emergency Procedure/Severe Accident Guidelines (EPGs/SAGs) for limiting Reactor Pressure Vessel (RPV) depressurization by reclosing Safety Relief Valves (SRVs) to conserve mass (aligns EALs with Emergency Operating Procedures (EOPs))

EP FAQ 2015-004, regarding Revision 3 of BWROG EPGs/SAGs and primary containment flooding (aligns EALs with EOPs)

EP FAQ 2015-005, regarding Revision 3 of BWROG EPG/SAG guidance for intentional venting of primary containment

EP FAQ 2015-006, clarifying the BWR wetwell as a "direct" release path

EP FAQ 2015-014, considering relevant operating modes for key safety functions

EP FAQ 2015-015, regarding a table listing of available power sources. 3.3 NRG Information Notice Evaluation NRG Information Notice (IN) 2013-01, "Emergency Action Level Thresholds Outside the Range of Radiation Monitors," (Reference

6) informs licensees of the importance of adequate controls to properly evaluate site configuration changes for impact on the ability to maintain an effective emergency plan. Specifically, the IN informs licensees of issues that arose when radiation monitors were not properly evaluated in conjunction with changes made to EAL thresholds for emergency classifications.

Similar issues are discussed in NRG IN 2005-19, "Effect of Plant Configuration Changes on the Emergency Plan," (Reference 7). NEI 99-01, Section 4.3, (Reference

1) also provides guidance for instrumentation used for EALs. NSPM's process for revision and control of EALs includes a validation and verification process that ensures the EALs can be used effectively as written and are compatible with plant conditions, hardware, and procedures.

The process confirms that the EAL is technically correct and is in conformance with design inputs and bases. NSPM has performed reviews per the described process and has concluded that the proposed EAL threshold values are within the calibrated ranges of the instrumentation referenced in the EALs and the threshold values can be accurately read. 3.4 Radiation Monitor Differences Between Current and Proposed EALs As part of the change to NEI 99-01, Revision 6, EALs, NSPM updated radiation monitor calculations and reviewed radiation monitors (RMs) for relevance to specific initiating conditions in the proposed EAL scheme. Based on that review, there is a difference between the radiation monitors used in the current EAL scheme and the proposed EAL scheme for EAL RS1 .1 and RG1 .1. Currently, the MNGP EAL RS1.1 for initiating condition (IC), "Off-site Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mRem TEDE [Total Effective Dose Equivalent]

or 500 mRem Thyroid COE [Committed Dose Equivalent]

for the Actual or Projected Duration of the Release" and RG1 .1 for IC, "Off-site Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mRem TEDE or 5000 mRem Thyroid Page 5 of 13 L-MT-17-012 Enclosure NSPM COE for the Actual or Projected Duration of the Release Using Actual Meteorology," both use a valid reading on the Reactor Building (RB) Vent Effluent Monitor (Channel A or B), along with other RMs, as indication of offsite doses exceeding the respective IC. The RB Vent Effluent Monitor will not be used in the proposed EALs RS1 .1 and RG1 .1 based on plant design as discussed below. Reactor Building (RB) Ventilation Exhaust Air Monitoring Subsystem Configuration:

The RB ventilation exhaust air monitoring subsystem measures the radioactivity in the combined exhaust from the Reactor Building, Radwaste Building, Turbine Building, Recombiner Building, and Chemistry Laboratory ventilation systems. Provision is made for indication and recording in the control room and for automatic alarm when radioactivity reaches prescribed levels. Two monitoring channels are provided for the RB Vent Exhaust Plenum, each consisting of a Geiger Mueller (GM) tube detector and an associated trip function.

The trip function isolates the normal RB ventilation exhaust path, initiates the Standby Gas Treatment (SBGT) System, and closes select primary containment isolation valves. The RM channels and the trip function are safety-related.

Additional monitoring of the RB ventilation exhaust subsystem is provided by the RB Vent Wide Range Gas Monitoring (RBV WRGM) subsystem.

The RBV WRGM subsystem consists of two trains of instrumentation each of which includes an isokinetic sampling probe assembly, a sample conditioning unit, a sample detection unit with noble gas activity sensors and sample sumps, and associated electronics.

The RBV WRGM is referred to in current Revision 4 MNGP EALs RU1, RA1, RS1 and RG1 as the RB Vent Effluent Monitor (Channel A or B). The RB Vent Effluent Monitoring system provides indication and recording of reactor building vent noble gas activity.

The sensitivity and range of the RB Vent Effluent Monitor is such that equipment is capable of detecting activity levels consistent with the Lower Limit of Detection (LLD) requirements as specified in the Offsite Dose Calculation Manual (ODCM). The monitor alarms in the control room upon detection of an elevated release rate in the RB vent. Control room operators can initiate prompt manual isolation of secondary containment and actuation of the SBGT system to terminate a release that could result in exceeding the limits of 10 CFR 50, Appendix I. The flow past the RB Vent Effluent Monitors is isolated when the RB Vent Exhaust Plenum RM isolates the normal RB ventilation exhaust path. As noted above, the RB Vent Exhaust Plenum RMs will provide a signal that isolates the normal RB ventilation exhaust path and initiates SBGT. These actions direct the gaseous radioactive effluents through the SBGT system to the off-gas stack. The effluent is then monitored by the Stack Effluent Monitors (Channels A and B). RB vent isolation occurs prior to core damage events, which are potential precursors to releases of radioactivity leading to higher emergency classifications (e.g., Site Area Emergency or General Emergency).

Page 6 of 13 L-MT-17-012 Enclosure NSPM The Stack Effluent Monitors (Channels A and B) monitor the radiation level of the gas sampled at the stack. The Stack Effluent Monitors consist of a wide range gas monitor having two channels, each with an isokinetic probe assembly, a sampling condition unit, a sample detection unit, and electronics for monitoring in the control room. If the observed radiation release level exceeds a preset level, the system provides a HIGH upscale alarm signal to the control room. If the radiation release level continues to increase to a second preset level, the system provides a HIGH-HIGH upscale alarm and trip signal to terminate stack releases from the air ejector and off-gas storage system. The Stack Effluent Monitor (Channel A or B) readings are also EAL entry points in the current EALs for RU1, RA1, RS1, and RG1. Changes in RMs for Proposed Revision 6 EAL RS1 and RG1: The current EAL thresholds for RU1, RA 1, RS1, and RG1 all use the RB Vent Effluent Monitor (Channel A or B). As discussed above, the MNGP design includes the RB Vent Exhaust Plenum Monitor that isolates and routes the normal RB ventilation through SBGT to the off-gas stack. The isolation automatically occurs when the RB Vent Exhaust Plenum Monitor reaches or exceeds 26 mrem/hr. This value corresponds to a Fuel Handling Accident release rate of 1.08E+06 µCi/sec. This release rate is below the EAL threshold for a Site Area Emergency (SAE) as calculated in CA-04-199 (see Attachment 4). Therefore, the RB Vent Exhaust Plenum Monitor will isolate the normal RB ventilation prior to the SAE threshold.

With the ventilation path isolated; EAL thresholds,for the RB Vent Effluent Monitor (Channel A or B) for a SAE and a General Emergency would not be valid (no flow). The RB Vent Effluent Monitor (Channel A or B) will continue to be used in the Unusual Event (UE) and Alert EAL thresholds.

3.5 EAL Scheme Change Conclusions 10 CFR 50, Appendix E, Section IV.B.2 stipulates that a licensee desiring to change its entire EAL scheme shall submit an application for an amendment to its license and receive NRG approval before implementing the change. The proposed change to adopt the NEI 99-01, Revision 6, EAL scheme has been reviewed by NSPM and determined to not reduce the effectiveness of the Emergency Plan to meet the applicable emergency planning requirements established in 10 CFR 50.47 and 10 CFR 50, Appendix E. Further, the proposed change will continue to provide consistent emergency classifications.

Accordingly, pursuant to the requirements of 1 O CFR 50, Appendix E,Section IV.B.2, NSPM requests NRG review and approval of the proposed changes to the MNGP EAL scheme in accordance with 10 CFR 50.90. 4.0 REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria The following are applicable to the proposed amendment:

Page 7 of 13 L-MT-17-012 Enclosure NSPM 10 CFR 50.47, "Emergency Plans", sets forth the NRG emergency plan requirements for nuclear power plant facilities.

Section (a)(1 )(i) of 10 CFR 50.47 states in part: No initial operating license for a nuclear power reactor will be issued unless a finding is made by the NRG that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency." 10 CFR 50.47(b) establishes the standards that onsite and offsite emergency response plans must meet for the NRG staff to make a positive finding that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.

The planning standard of 10 CFR 50.47(b)(4) requires that onsite and offsite emergency response plans contain the following:

A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures.

10 CFR 50, Appendix E, Section IV, "Content of Emergency Plans," Item B, "Assessment*Actions," stipulates that Emergency Plans include EALs.Section IV.B.1 of Appendix Estates: The means to be used for determining the magnitude of, and for continually assessing the impact of, the release of radioactive materials shall be described, including emergency action levels that are to be used as criteria for determining the need for notification and participation of local and State agencies, the Commission, and other Federal agencies, and the emergency action levels that are to be used for determining when and what type of protective measures should be considered within and outside the site boundary to protect health and safety. The emergency action levels shall be based on in-plant conditions and instrumentation in addition to onsite and offsite monitoring.

By June 20, 2012, for nuclear power reactor licensees, these action levels must include hostile action that may adversely affect the nuclear power plant. The initial emergency action levels shall be discussed and agreed on by the applicant or licensee and state and local governmental authorities, and approved by the NRG. Thereafter, emergency action levels shall be reviewed with the State and local governmental authorities on an annual basis. The NRG requires that licensees adopting a new EAL scheme submit a license amendment request for approval in accordance with 1 O CFR 50, Appendix E,Section IV.B.2, which states: Page 8 of 13 L-MT-17-012 Enclosure NSPM A licensee desiring to change its entire emergency action level scheme shall submit an application for an amendment to its license and receive NRG approval before implementing the change. Licensees shall follow the change process in § 50.54(q) for all other emergency action level changes. NSPM evaluated the proposed change and determined it maintains conformance with the regulatory requirements described above. Regulatory Guidance:

NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors," Revision 6 (Reference 1 ), provides guidance to nuclear power plant operators for the development of a site-specific emergency classification scheme. The methodology described in NEI 99-01, Revision 6, is consistent with Federal Regulations and related US NRG requirements and guidance.

The NEI 99-01, Revision 6, methodology has been endorsed by the NRG (Reference

2) as an acceptable approach to meeting the requirements of 10 CFR 50.47(b)(4) and of 10 CFR 50, Appendix E. Additionally, several guidance documents were consulted during the development of this license amendment request These include:

RG 1 :219, "Guidance On Making Changes to Emergency Plans for Nuclear Power Plants," Revision 1 (Reference

11)

NRG Information Notice (IN) 2013-01, "Emergency Action Level Thresholds Outside the Range of Radiation Monitors," dated February 13, 2013 (ADAMS Accession No. ML 12325A326) (Reference

6)

NRG Information Notice 2005-19, "Effect of Plant Configuration Changes on the Emergency Plan," dated July 18, 2005 (ADAMS Accession No. ML051530520) (Reference

7)

RIS 2003-18, "Use of NEI 99-01, 'Methodology for Development of Emergency Action Levels,' Revision 4, Dated January 2003," (Reference

8) and its supplements (References 9 and 10)

EP FAQ 2015-003, regarding Revision 3 of Boiling Water Reactor Owners Group (BWROG) Emergency Procedure/Severe Accident Guidelines (EPGs/SAGs) for limiting RPV depressurization by reclosing SRVs to conserve mass (aligns EALs with Emergency Operating Procedures (EOPs))

EP FAQ 2015-004, regarding Revision 3 of BWROG EPGs/SAGs and primary containment flooding (aligns EALs with EOPs)

EP FAQ 2015-005, regarding Revision 3 of BWROG EPG/SAG guidance for intentional venting of primary containment

EP FAQ 2015-006, clarifying the BWR wetwell as a "direct" release path

EP FAQ 2015-014, considering relevant operating modes for key safety functions

EP FAQ 2015-015, regarding a table listing of available power sources. Page 9 of 13 L-MT-17-012 Enclosure 4.2 Precedent NSPM This request is similar to approved requests to adopt the NEI 99-01, Revision 6, scheme change LARs for the following licensees:

V.C. Summer Nuclear Station Unit 1 (ADAMS Accession No. ML 15063A355)

Exelon Generation Company (ADAMS Accession No. ML 15141 A058) -Braidwood Station, Units 1 and 2 -Byron Station, Units 1 and 2 -Clinton Power Station, Unit 1 -Dresden Nuclear Power Station, Units 1 , 2 and 3 -LaSalle County Station, Units 1 and 2 -Limerick Generating Station, Units 1 and 2 -Oyster Creek Nuclear Generating Station -Peach Bottom Atomic Power Station, Units 1, 2 and 3 -Quad Cities Nuclear Power Station, Units 1 and 2 -Three Mile Island Nuclear Station, Units 1 and 2

Catawba Nuclear Station, Units 1 and 2 (ADAMS Accession No. ML 16082A038) 4.3 No Significant Hazards Consideration Determination In accordance with the requirements of 1 O CFR 50.90, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), requests changes to the Emergency Plan for the Monticello Nuclear Generating Plant (MNGP) to adopt the NRG-endorsed Nuclear Energy Institute (NEI) 99-01, Revision 6, emergency action level (EAL) scheme. The proposed change does not reduce the effectiveness of the Emergency Plan to meet the emergency planning requirements established in 10 CFR 50.47 and 10 CFR 50, Appendix E. NSPM has evaluated the proposed amendment against the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," and has determined that the proposed amendment presents no significant hazards. NSPM's evaluation against each of the criteria in 10 CFR 50.92 follows. 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:

No. The proposed change to the MNGP EAL scheme does not impact the physical function of plant structures, systems or components (SSC) or the manner in which the SSCs perform their design function.

The proposed change neither adversely affects accident initiators or precursors, nor alters design assumptions.

Therefore, the proposed change does not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an event. The Page 10 of 13 L-MT-17-012 Enclosure NSPM Emergency Plan, including the associated EALs, is implemented when an event occurs and cannot increase the probability of an accident.

Further, the proposed change does not reduce the effectiveness of the Emergency Plan to meet the emergency planning requirements established in 10 CFR 50.47 and 1 O CFR 50, Appendix E. Therefore, the proposed EAL scheme change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response:

No. The proposed change does not involve any physical alteration to the plant, that is, no new or different type of equipment will be installed.

The proposed change also does not change the method of plant operation and does not alter assumptions made in the safety analysis.

Therefore, the proposed change will not create new failure modes or mechanisms that could result in a new or different kind of accident.

The Emergency Plan, including the associated EAL scheme, is implemented when an event occurs and is not an accident initiator.

Therefore, the proposed EAL scheme change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety? Response:

No. Margin of safety is provided by the ability of accident mitigation SCCs to perform at their analyzed capability.

The change proposed in this license amendment request does not modify any plant equipment and there is no impact to the capability of the equipment to perform its intended accident mitigation function.

The proposed change does not impact operation of the plant or its response to transients or accidents.

Additionally, the proposed changes will not change any criteria used to establish safety limits or any safety system settings.

The applicable requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E will continue to be met. Therefore, the proposed EAL scheme change does not involve a significant reduction in a margin of safety. Based on the above evaluation, the NSPM concludes that the proposed change does not involve a significant hazards consideration as defined in 10 CFR 50.92(c), and accordingly, a finding of no significant hazards consideration is justified.

Page 11 of 13

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L-MT-17-012 Enclosure 4.4 Conclusions NSPM In conclusion, based on the considerations discussed in above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. 5.0 ENVIRONMENTAL CONSIDERATION NSPM has determined that the proposed change would not revise a requirement with respect to installation or use of a facility or component located within the restricted area, as defined in 1 O CFR 20, nor would it change an inspection or surveillance requirement.

The proposed amendment does not involve (i) a significant hazards consideration, or (ii) authorize a significant change in the types or a significant increase in the amounts of any effluent that may be released offsite, or (iii) result in a significant increase in individu.al or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for a categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, NSPM concludes that pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Nuclear Energy Institute (NEI) 99-01, Revision 6, "Methodology for Development of Emergency Action levels for Non Passive Reactors," November 2012 (ADAMS Accession No. ML 12326A805)

2. Letter from Mark Thaggard (US NRG) to Susan Perkins-Grew (Nuclear Energy Institute), "U.S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 6, November 2012 (TAC NO. D92368)," dated March 28, 2013 (ADAMS Accession No. ML 12346A463)

3. Letter from US NRG to J Conway, "Monticello Nuclear Generating Plant, Unit No. 1 -Revision To Emergency Plan Emergency Action Levels (TAC No. MC5017)," dated January 5, 2006 (ADAMS Accession No. ML060040437)

4. Letter from T.J Palmisano (NMC) to NRG, L-MT-04-066, "Revision to Emergency Action Levels," dated October 22, 2004 (ADAMS Accession No. ML053420669)

Page 12 of 13 L-MT-17-012 Enclosure NSPM 5. NRG Bulletin 2005-02, "Emergency Response Preparedness and Response Actions for Security-Based Events," issued July 18, 2005 (ADAMS Accession No. ML051740058)

6. NRG Information Notice (IN) 2013-01, "Emergency Action Level Thresholds Outside the Range of Radiation Monitors," dated February 13, 2013 (ADAMS Accession No. ML 12325A326)

7. NRG Information Notice 2005-19, "Effect of Plant Configuration Changes on the Emergency Plan," dated July 18, 2005 (ADAMS Accession No. ML051530520)

8. Regulatory Information Summary (RIS) 2003-18, "Use of NEI 99-01, 'Methodology for Development of Emergency Action Levels,' Revision 4 Dated January 2003," dated October 8, 2003 (ADAMS Accession No. ML032580518)

9. Regulatory Information Summary (RIS) 2003-18, Supplement 1, "Use of NEI 99-01, 'Methodology for Development of Emergency Action Levels,' Revision 4, Dated January 2003," dated July 13, 2004 (ADAMS Accession No. ML041550395)

10.

Regulatory Information Summary (RIS) 2003-18, Supplement 2, "Use of NEI 99-01, 'Methodology for Development of Emergency Action Levels,' Revision 4, Dated January 2003," dated December 12, 2005 (ADAMS Accession No. ML051450482)

11. Regulatory Guide (RG) 1.219, "Guidance On Making Changes to Emergency Plans for Nuclear Power Plants," Revision 1, dated July 2016 (ADAMS Accession No. ML16061A104)

Page 13 of 13 L-MT-17-012 NSPM ATTACHMENT 1 MONTICELLO NUCLEAR GENERATING PLANT License Amendment Request to Revise the Emergency Action Level Scheme Emergency Action Level Comparison Matrix Document (Deviations and Differences)

(73 pages to follow)

NEI 99-01 Rev 6 Deviations and Differences Monticello Nuclear Generating Plant Table of Contents Generic Differences

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I HG I: Initiating Conditions

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3 S RGI: Initiating Conditions

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2 HG7: Initiating Conditions

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39 RG2: Initiating Conditions

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3 HS I: Initiating Conditions

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40 RS I : Initiating Conditions

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.4 HS6: Initiating Conditions

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4I RS2: Initiating Conditions

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5 HS7: Initiating Conditions

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42 RAI: Initiating Conditions

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6 HAI: Initiating Conditions

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43 RA2: Initiating Conditions

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S HA5: Initiating Conditions

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44 RA3: Initiating Conditions

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I 0 HA6: Initiating Conditions

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45 RUI: Initiating Conditions

......................................................................... I 2 HA7: Initiating Conditions

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46 RU2: Initiating Conditions

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I4 HUI: Initiating Conditions

....................*................................................... 4 7 CGI: Initiating Conditions

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I5 HU2: Initiating Conditions

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4S CS I: Initiating Conditions

.......................................................................... I 7 HU3: Initiating Conditions

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49 CAI: Initiating Conditions

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.... I9 HU4: Initiating Conditions

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50 CA2: Initiating Conditions

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20 HU7: Initiating Conditions

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52 CA3: Initiating Conditions

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21 SG I: Initiating Conditions

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53 CA6: Initiating Conditions

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23 SGS: Initiating Conditions

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54 CUI: Initiating Conditions

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25 SSI: Initiating Conditions

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55 CU2: Initiating Conditions

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26 SS5: Initiating Conditions

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56 CU3: Initiating Conditions ......................................................................... 27 SSS: Initiating Conditions

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5S CU4: Initiating Conditions

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2S SAI: Initiating Conditions

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59 CU5: Initiating Conditions ......................................................................... 29 SA2: Initiating Conditions

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60 EUI: Initiating Conditions

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.30 SA5: Initiating Conditions

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62 BWR Fission Product Barrier Matrix -Initiating Conditions/Thresholds

.. .3 I SA9: Initiating Conditions

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63 . I. RCS Activity .........................................................................

.3 I SUI: Initiating Conditions

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65 2. RPV Water Level .... : .. ************************************************************32 SU2: Initiating Gonditions

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66 3. RCS Leak Rate ......................................................................

33 SU3: Initiating Conditions

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67 4. Primary Containment Radiation

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35 SU4: Initiating Conditions

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6S 5. Other Indications

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36 SU5: Initiating Conditions

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69 6. Emergency Director Judgment ..............................................

36 SU6: Initiating Conditions

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7I GENERIC DIFFERE N CES NE I 99-0 1 R ev 6 -M ontic e llo N ucl e a r G e nerating P l a n t c References P WRs Deleted PWR references as aoorop ria te Use s A for the radiological effluent/radiation level ICs U s es R for the radiolocical effluent/radiation level IC s Uses E-HU for the ISFSI ICs Uses EU for the ISFSI ICs Emergency C l assification ICs are presented in ascending order (NOUE -GE) Emergency Classification ICs are presented in descending order (GE -NOUE) EAL Thresholds are i n dicated in numeric format-(l), (2), etc. EAL Thresho l ds include I C designator combined with the applicab l e EAL T h reshold number (e.g., RGl.1, RGl.2, CGl.l, CSJ .1, etc.) Includes Hot Standby as an Operating Mode Deleted Hot Standby as an Operating Mode -not u sed by BWRs BWROG EPG/SAG Revision 3 guidance not included in FPB Matrix and Incorporated EP FAQ 2015-003, 2015-004 , and 2015-005 guidance in EALs Basis and/or Basis as applicab l e 'Direct' release to the environment is not defined in FPB Matrix Basis Incorporated EP FAQ 2015-006 guidance in FPB Matrix Basis Does not include site specific basis references S i te specific references are incl u ded in the basis Uses generic Mode De sc riptions Uses MNGP Technical Specification Mode Descriptions Uses 'increase' i n the following ICs/EAL thre s holds: CGl; CS!; CAI; CA3; 'Rise' is u sed instead of 'increa se' consistent with s tandard communications CUl; and, CU3. terminology. The words 'increase' and 'decrease' are not used because the y are easily misunderstood. 'Rise' i s equivalent to 'increase'.

GENERAL NOTES Instrument setpoi n t readings used as threshold val u es to determine emergency classifications have been verified by Monticello personne l as being within the range of the instr u ment and clearly and consistently read within the scale of the instrument.

This va l id a tion is performed IA W FP-R-EP-05 , Revision and Control of Emergency Action Levels (Technical Bases and Matrix), and documented on Form QF-0744 , EAL Verification and Validation.

Site s pecific information is highlighte dl in yellow. ODCM is the contro ll ing Radio l ogical Effluent Doc u ment. EPFAQ 2015-008 guidance was considered and determined to be not applicab l e to MNGP since alternate power sources are not permanently installed and if used would only supp l y power to specific loads and not the full l oad of the emergency buses. EPFAQ 2015-013 guidance was reviewed and considered for implementation.

No revisions to HG 1 were made as a result of this review. EPFAQ 2015-014 g u idance related to Mode APolicability incorporated in HS6. EPFAQ 2015-015 guidance used to deve l op Tab l e Sl and Table S2. Rep l aced ANY with E I THE R w h enever two EAL thresholds are identified.

Revised Section 5.3 examples to reflect single unit applicability for MNGP. Added site specific clarifying informat i on to the bas i s for CUl. Added site specific clarifying information to the basis for CU4, SG8, and SS8. Ad d ed site specific clarifying informat i on to the bas i s for CA2. Added site specific clarifying information related to the OBE alarm setpoint to the basis for HU2. Revised basis examples for CU2 and SA l to reflect site specific information.

Aooe n dix A -D e l eted PWR Acronyms and Abbreviations. Added site specific acronyms and abbreviations as appropriate.

Appendix B -Incorporated Site Specific definit i o n s as appropriate.

1 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFL U ENT ICS/EALS RGl: INITIATING CONDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Rele ase of gaseous radioactivity resulting in offsite d ose greater than 1 , 000 Rele ase of gaseous radioactivity resulting in offsite dose greater than 1,000 rnrem TEDE or 5 , 000 rnrem thyroid CDE. rnrem TEDE or 5 , 000 mrem thyroid CDE. Difference I Deviation I Justification None THRESHOLDS NEI 99-01Rev6 Monticello Nuclear Generating Plant (1) Reading on ANY of the following radiation monitors greate r than RG l.l Re adi n g on the following radiation monitor greater th an the reading t he re ad ing s hown for 15 minutes or longer: s hown for 15 minute s or lon ger: (s ite-specific monitor li st and threshold val ue s) I Stack Effluent Monitor (Ch A or B) I 8 E+07 µCi/s e q I (2) Do se assessment u s in g actual mete oro logy indi cates doses greater RGl.2 Do se assessment usin g actual meteorology indicates dose s grea ter than 1,000 rnrem TEDE or 5,000 rnrem thyroid CDE at or beyond than 1 , 000 mrem TEDE or 5,000 mrem thyroid CDE at or be yo nd (site-specific do se receptor point). the SITE (3) Field survey r es ult s indicate EITHER of the fo ll owing at or beyond RGl.3 Field s ur vey results indic ate EITHER of the following at or beyond (s ite-specific do se receptor point): t h e S IT E B OUNDA R Y:

Closed window dose rates greater th a n 1 , 000 mR/hr expected to

Closed window do se rates greater th a n 1,000 mR/hr expec t e d to continue for 60 minute s or longer. continue for 60 minutes or l o n ger.

Analyses of field s urvey samp l es indicate thyroid CDE greater

Analyses of field s urvey sa mple s indicate thyroid CDE g reater than 5,000 mrem for one hour of inhal a tion. than 5,000 mrem for one hour of inhal a tion. Difference I Deviation I Justification Difference:

Site specific information provided.

Justification:

Calculation 04-199 provides the basis for the Stack Effluent Monitor reading related to RGl.1. The MNGP ODCM (Figure 1) identifies the site boundary as the site-specific dose receptor point. 2 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUEN T ICS/EALS RG2: IN I TIATING CONDITIONS NEI 99-01 Rev 6 Mo n ticello Nuclear Generating Plant Spent fuel pool le vel ca nn ot be restored to at l east (si t e-specific Level 3 Spent fuel pool l evel cannot be restored to at l eas t 15.25' for 60 minutes or d esc r ip ti o n) fo r 60 minutes or longer. l o n ger. Difference I Deviation I Justification Difference:

Site specific information provided for Level 3 indication.

Justification:

A level indication of 15.25' corresponds to the Level 3 value for MNGP. Although this value is slightly above the top of the fuel, it is the lowest value that the installed instrumentation will read. Indication is provided in the control room with a dedicated instrument located in the control room envelope.

THRESHOLDS

-NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) Spent fuel pool l eve l can n ot be restore d to a t least (s i te-specific RG 2.l Spent fuel pool le ve l can n ot be restored to at least 15.2 5' for 60 Level 3 va lu e) for 60 minutes or longer. minutes or l onger. Difference I Devia t ion I Justification Difference

Site specific information provided for Level 3 indication.

Justification:

A level indication of 15.25' corresponds to the Level 3 value for MNGP. Although this value is slightly above the top of the fuel, it is the lowest value that the installed instrumentation will read. Indication is provided in the control room with a dedicated instrument located in the control room envelope.

3 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS RSl: INITIATING CONDITI O NS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Release of gaseous radioactivity resulting in offsite dose greater than 100 Release of gaseous radioactivity resulting in off s ite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE. nu*em TEDE or 500 mrem thyroid CDE. Difference I Deviation I Justification None THRESH O L D S NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) Reading on ANY of the following radiation mom tors greater than RSl.l Reading on the following radia t ion monitor greater than the reading the reading shown for 15 minutes or longer: shown for 15 minutes or longer: (si t e-specific monitor list and threshold values) I St a ck Effluent Monitor (Ch A or B) I 8 E+06 µ Ci/se d I (2) D ose assessment using actual meteorology indicates doses greater RSl.2 Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond (site-than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond th e specific dose receptor point). S ITE BOUNDAR (3) Field survey results indicate EITHER of the fo ll owing at or beyond RSl.3 Field survey resu lt s 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

Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer. continue for 60 rrunutes or longer.

Ana l yses of field survey samples indicate thyroid CDE greater

Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation. than 500 mrem for one hour of inhalation.

Difference I Deviation I Justification Difference:

Site specific information provided.

Justification:

Calculation 04-199 provides the basis for the Stack Effluent Monitor reading related to RSl.1. The MNGP ODCM (Figure 1) identifies the site boundary as the site-specific dose receptor point. 4 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS R S2: INITIATING CO N D ITI O NS NEI 99-01 Rev 6 Mont i cello N uclear Generati ng P l ant Spent fuel pool l evel at (site-specific Leve l 3 description).

Spe n t fue l pool l evel at 15.25'. Difference I De v iation I Justification Difference:

Site s pecific info r mation pro v ided for L evel 3 indication.

Justification:

A level indication of 15.25' corresponds to the Level 3 value for MNGP. Although this value is slightly above tl1e top of the fuel, it is the lowest value that the installed instrumentation will read. Indication is provided in the control room with a dedicated instrument located in the control room envelope.

THRESHOL D S NEI 99-01 Re v 6 Montice ll o N uclear Generatin g Plant (1) Lowe rin g of s p e n t fue l pool l eve l to (site-specific Leve l 3 va lu e). R S2.l L oweri n g of s p e n t fue l pool l eve l to 15.25'. Difference I De v iation I Justification Difference:

Site s pecific information pro v ided for Le vel 3 i ndication.

Justification:

A level indication of 15.25' corresponds to the Level 3 value for MNGP. Although this value is slightly above the top of the fuel, it is the lowest value that the installed instrumentation will read. Indication is provided in the control room with a dedicated instrument located in the control room envelope.

5 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS RAl: INITIAT I NG C O NDITI O NS NEI 99-01 Re v 6 Monticello Nuclear Generating Plant Release of gaseous or liquid radioactivity resulting in offsite do se greater Relea se of gaseous or liquid radioactivity resulting in offsite do se greater than 10 mrem TEDE or 50 mrem thyroid CDE. than 10 mrem TEDE or 50 mrem thyroid CDE. Difference I Deviation I Justification None TH RE SH O L D S NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) Reading on ANY of the following radiation monitors greater than RAl.l Re ad ing on ANY of the following radiation monitors greater than the reading s hown for 15 minutes or longer: the reading s hown for 15 minutes or longer: (site-specific monitor list and threshold values) Gaseous Efflue n t Monitors (2) Dose assessment u s in g actual meteorology indic ates doses greater Stack Effluent Monitor (Ch A or B) 8 E+05 µCi/sec than 10 mrem TEDE or 50 mrem th yroi d CDE at or beyond (s ite-,RB Vent Effluent Monitor (Ch A or B) 6 E+05 µCi/s ec specific dose receptor point). Liquid Effluent Monitors (3) A n alysis of a liquid efflue nt sample indicates a concentration or p i sc h arge Ca n a l 2000 cps release rate that would result in dose greater th a n 10 mrem TEDE Se r vice W ater f?OO cps or 50 mrem thyroid CDE at or beyond (site-s pecifi c do se receptor if BNWS 9 E+04 cp nil point) for one hour of exposure.

RAl.2 Do se assessment u s ing actual meteorology indicate s do ses g reat er (4) Fie ld survey re s ults indicate EITHER of the following at or beyond than J 0 mrem TEDE or 50 mrem thyroid CDE at or beyond the (site-s pecific do e receptor point): SITE BOUNDAR'v.

Clo se d wi ndow dose rates greater than 10 mR/hr expected to RAl.3 Analysis of a liquid effluent sample indicates a concentration or continue for 60 minute s or longer. release rate that would result in do ses greater than 10 mrem TEDE

Analyse s of field s urvey samples indicate thyroid CDE greater or 50 mrem thyroid CDE at or beyond the SIT E B OUN D ARY for than 50 mrem for one hour of inhalation. one hour of exposure.

RAl.4 Fi e ld survey results indicate EITHER of the following at or b eyo nd the S I TE B OUN D A R

Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or l onger.

Analyses of field survey sam pl es indicate thyroid CDE grea t er than 50 mrem for one hour of inhalation.

6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS Difference I Deviation I Justification

. Difference:

Site specific information provided.

Justification:

Calculation 04-199 provides the basis for the Effluent Monitors and readings related to RAl.1. The MNGP ODCM (Figure 1) identifies the site boundary as the site-specific dose receptor point. 7 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS RA2: INITIATING C O NDITIONS NEI 99-01 Rev 6 Mo n ticello Nuclear Generating Plant Significant l owering of water level above , or damage to, inadiated fuel. Significant lowering of water level a bove , or damage to , inadiated fuel. Difference I Deviation I Justification None THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant ( 1) Uncovery of irradia t ed fuel in the REFUELING PATHWAY. RA2.1 Uncovery of irradiated fuel in the REFUELING PATHWAY. (2) D a mage to irradiated fuel resulting in a release of radioactivity from RA2.2 D a mage to inadiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of the following radiation monitors:

the fuel as indicated by ANY of the following radiation monitors: (site-specific li st ing of rad i a tion monitors , and the associated Monitor Alaq;n/Trip readings , setpoints and/or alarms) A-1 1027 RB NE Low 20 mR/hr (3) Lowering of spent fuel pool le v el to (s ite-s pecific Level 2 value). A-2 1027 RB N Hig l] QOO mR/hr A-3 1027 RB W Stairwa y 100 mR/hr RM-l 7-452A Reactor Bu" ldir;i g Venti!a t i22J Q6 mRfhr Exhaust Plenum Monitor Ch A RM-l 7-452B Reactor Buil ding Ventilation 26 mR/hi Ex h aust Plenum Monitor Ch B RM-17-453A Fuel Pool Radiation Monitor Ch A 50 rnR/hr ,RM-17-453B Fuel Pool Radiatior;i Monitor Ch B [S o rnR/hr RA2.3 Lowering of spen t fuel pool level to 24.75'. Difference I Deviation I Justification 8

ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS Difference:

MNGP RA2.1 specifies the Refueling Pathway as the reactor refueling cavity, spent fuel pool, or fuel transfer canal. Site specific information provided for RA2.2. Justification:

Alarmffrip setpoints are in annunciator response procedures C.6-004-A-01, C.6-005-A-Ol, and C.6-005-A.02.

Control Room personnel are provided with guidance for response to alarms related to A-1, A-2, and A-3 in annunciator response procedure C.6-004-A-Ol.

Difference:

Site specific information provided for RA2.3. Justification:

A level indication of24.75' corresponds to the Level 2 value for MNGP. Indication is provided in the control room with a dedicated instrument located in the control room envelope.

9 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUEN T ICS/EALS RA3: INITIATING CONDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Radiation levels that impede access to equipment nece ssary for normal plant Radiation levels that impede access to equipment neces sary for norm a l plant operations , cooldown or s hutdown. operations, cooldown or shutdown.

Difference I Deviation I Justification None THRESH O LDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) Dose rate greater than 15 mR/hr in ANY of the following areas: RA3.l Dose rate greater than l 5 mR/hr in ANY of the following areas:

Control Room

Control Room (A-20 Control Roo rn Low Range)

Central Alarm Station

Centr a l Alarm St atio n (by s ur vey) * (o ther site-specific areas/rooms)

Secondary A larlII St a tion (by survey) (2) An UN PLANN E D eve nt results in radiation levels that prohibit or RA3.2 An UN PLANNED eve nt results i n radiatio n levels t h at prohibit or impede ac ce ss to any of the fo ll owing plant rooms or areas: impede access to any of the Table HI pl a nt rooms or areas: (s ite-s pecific list of plant rooms or areas with entry-related mode !fable Hl appl ic abiuty identified) Building Rooms App li cable Mode(s) ,Reactor Build i n g All All T u r b i n e Bu i ldin g All A l l Intake Structure All Ail1 Difference I Deviation I Justification Difference:

NEI 99-01 Rev 6 Threshold (1) does not specify specific radiation monitors or methods to determine dose rates. MNGP Threshold RA3.1 specifies applicable rad monitor for the control room and methods to determine dose rates at other identified areas (Central A l arm Station and Secondary Alarm Station). Ops Man B.05.12-06 identifies the range for ARM A-20 used in RA3.1. Site specific information prov i ded. Justification:

Human Factors consideration

-identification of the applicable control room radiation monitor facilitates the determination by site personnel that the threshold has been met. Radiation monitors are not available in the Central or Secondary Alarm Stations; therefore, local surveys are required to determine if the threshold has been met. 10 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS Difference:

NEI 99-01 Rev 6 Threshold (2) does not include table reference.

MNGP Threshold RA3.2 includes reference to Table Hl and incorporates the table. RA3.2 also provides site specific plant locations which require access during normal operating conditions during the specified modes. Justification:

Human Factors consideration

-use of table format clearly identifies the applicable rooms and plant modes. 11 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUE N T I C S/EALS RUl: INITIATING CONDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant R e le ase of gaseous or liquid ra dioacti vity greater than 2 time s the (s ite-Relea s e of gaseous or liquid radioactivity greater than 2 time s the ODC s pecific effluent re l ease controlli n g document) limit s for 60 minute s o r l imits for 60 minutes or l o n ge r. longer. Difference I Deviation I Justification None THRESH O L D S NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (l) Re ading on ANY effluent radia tion monitor greater than 2 times the RUl.l R eading on ANY of the following effluent radiation monitors (s ite-specific effluent release controlling document) limit s fo r 60 greater th a n the values for 60 minute s or lon ge r: minutes or lon ge r: Gaseous Eftluept Monitors (s ite-s pecific monitor li st and thre s h o l d va lue s correspo ndin g t o 2 Stael< Effluent Monitor (Ch A or B) 4 E+OS µCi/sec time s the controllin g document limits) IR1fVent Effluent Monitor (C!;l A or B) G E+04 µCi/sec (2) Reading o n ANY effl u e nt ra di a tion monitor greater t h a n 2 time s the [L iquid Effluent Monitors alarm set p oi nt estab l is h ed by a current radioactivity di charge Di scharge Cana] 900 cps permit for 60 minutes or l o n ger. Service Water GOO cps (3) Sample analy s i s for a gaseo u s or liquid release indic a tes a TBNWS 4 E+04 cpm concentration or relea se rate greater than 2 time s the (s it e-s pecific RUl.2 Re a din g on ANY effluent radiatio n monitor g re a ter than 2 time s the effl uent r e l ease co nt ro llin g document) limit s for 60 minutes or alarm s etpoint establis h ed by a curre nt radioactivity di sc h arge l o n ger. p er mit for 60 minute s or l onger. R Ul.3 Sample analysis for a gaseo us o r liquid rel ease indicates a co ncentration or re l ease ra t e grea t er than 2 times the ODCM li mits for 60 minutes or longer. Difference I Deviation I Justification Difference:

Site specific information provided for ODCM identified effluent points. The Stack Effluent Monitor value is set at the ODCM va l ue. Justification: Calculation CA-04-199 provides the basis for the identified rad monitor values. The calculation determined that the Stack Effluent Monitor NUE and Alert set points are virtually identical due to the isotopic mix during normal operation vs. LOCA. Therefore, the 12 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS NUE set point is established at the ODCM limit rather than the NEI guidance of two times this value. 13 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFL U ENT ICS/EALS RU2: IN I TIATING C O NDITI O NS NEI 99-01 Rev 6 Monticello Nuclear Generating Pl a nt UNPLANNED lo ss of water level above irradiated fuel. UNPLANNED lo ss of water level a bo ve irradiated fuel. Difference I Deviation I Justification None THRESHOL D S NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) a. UNPLANNED water l eve l drop in the REFUELING RU2.l a. UNPLANNED water l evel drop in the REFUELING PATHWAY as indicated by ANY of the followi n g: PATHWAY as indicated b y ANY of the following: (site-specific level indication s). Spent Fuel Pool low water level alaq n AND Visual observation of an u ncoptrolled water le vel drop b. UNPLANNED rise in area radiat ion levels as indicated by below a fuel pool s!ci mmer surge tank inlen ANY of the following radiation monitor s. Observatio n of water leakage into tbe drywell or t he (site-specific list of area radiation monitor s) eactor building rorn piping pe etrations surround i ng the drywell. AND b. UNPLANNED rise in area radiation le ve ls as indicated by ANY of the following radiation monit ors. A-1 10 27 RB NE Low A-2 I 027 RB N High! A-3 1027 RB W Stairwa y RM-l 7-45 3A or B Fue l Pool Monitoring S y stem Difference I Deviation I Justification Difference:

Si t e specific information provided.

Justification:

MNGP Annunciator Response Procedures (ARPs) C.6-004-B-33 and C.6-065-A-06 provide control room personnel with information related to EAL thresholds RU2.la and RU2.lb. 14 COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS CGl: INITIA TI NG C O N D ITIONS NEI 99-01 R ev 6 Montice ll o Nuclear Generating Plant Loss o f (reac t or vessel/RCS

[PWR] or RPV [BWR]) inventory affecting fuel Loss of RPV inventory affecting fuel clad integrity with containment clad integrity with containment challenged.

challenged.

Difference I Deviation I Justification None THRES HO L D S NEI 99-01Rev6 Monticello Nuclear Generating P l ant (1) a. (Reactor vessel/RCS

[PWR] or RPV [BWR]) level less than CGl.l a. RPV level less than -l 26 in. (T AF) for 30 minutes or longer. (site-specific level) for 30 minutes or longer. AND AN D b. ANY indication from the Containment Challenge Table Cl. b. ANY indication from the Containment Challenge Table CGl.2 a. RPV level cannot be monitored for 30 minutes or longer. (see below). AND (2) a. (Reactor vessel/RCS

[PWR] or RPV [BWR]) level cannot b. Core uncovery is indicated by EITHER of the following:

be monitored for 30 minutes or longer.

Refueling Floo r radiation monitor reading greater than AND R/hr b. Core uncovery is indicated by ANY of the following:

UNPLANNED rise in drywell floor or equipment drai * (Site-specific radiation monitor) reading greater than s ump levels of sufficien t magnitude to indicate core (site-specific value) uncovery

Erratic source range monitor indication

[PWR] AND

UNPLANNED increase in (site-specific sump and/or c. ANY indication from the Containment Challenge Table Cl tank) levels of sufficient magnitude to indicate core Containment Challenge Table Cl uncovery

SECONDARY CONTAINMENT not established*

(Other site-specific indications)

Greater than or equal to 6% hydrogen and greater tha n or equal to 5% oxyge in prima y containment AND

UNPLANNED rise in conta inm ent pressure o c. ANY indication from the Containment Challenge Table greater than 1.84 psig (see below).

Two or more Reactor Building areas exceed Ma Safe Radiation Leve l s (C.5-1300 , Table X) *If SECONDARY CONTAINMENT is re-established prior 15 COLD SHUTDOWN/

REFUELING SYSTEM MALFUNCTION ICS/EALS Containment Table

CONTAINMENT CLOSURE not established*

(Explosive mixture) exists inside containment

UNPLANNED increase in containment pressure

Secondary containment radiation monitor reading above specific value) [BWR] *If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

Difference I Deviation I Justification to exceeding the 30 minute time limit, then declaration of a General Emergency is not required.

Difference:

NEI 99-01 Rev 6 Threshold (2)b, last bullet, refers to Other site-specific indications of core uncovery.

No Other site-specific indications are used at MNGP. Justification:

MNGP does not use any other site-specific indications of core uncovery other than those already identified in Threshold CGl.2.b. Difference:

Table designator Cl assigned to Containment Challenge Table. Justification:

Human Factors consideration

-editorial change to clearly identify tables within the document.

Difference:

NEI 99-01 Rev 6 Containment Challenge Table specifies CONTAINMENT CLOSURE. MNGP Containment Challenge Table specifies SECONDARY CONTAINMENT.

Justification:

Loss of SECONDARY CONTAINMENT is equivalent to CONTAINMENT CLOSURE not being established.

Intent of NEI 99-01 Rev 6 is still satisfied.

Difference:

NEI 99-01 Rev 6 Containment Challenge Table does not specify a site specific value for indications of an UNPLANNED increase in containment pressure.

MNGP Containment Challenge Table includes a specific pressure value (1.84 psig). Justification:

Human Factors consideration

-use of a specific value provides site personnel with a clear indicator of when this threshold of the Containment Challenge Table is met. Difference:

Site specific information provided regarding RPV level, drywell floor and equipment drain sumps, explosive mixture, containment pressure, and radiation monitors.

Justification:

Information is identified in MNGP EOPs (C.5.1-1100, C.5.1-1200, and C.5.1-1300) and ARP C.6-065-A-06.

CA-04-202 provides the calculated Refueling Floor radiation monitor value. 16 COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCT I O N I CS/EALS CSl: INITIATING C O NDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory affecting core Loss of RPV inventory affecting core decay heat removal capability.

decay heat removal capa b i lit y. D i fference I Deviat i on I Justification None THRESH O LDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) a. CONTAINMENT CLOSURE not established.

CSl.l a. SECONDARY CONTAINMEN 1i not established.

AND AN D b. (R eactor vesse l/RCS [PWR] or RPV [BW R]) level less tha r. b. RPV level less than -47 in. (site-spec i fie level). CSL.2 a. SECONDARY CONTAINMEN esta bli shed. (2) a. CONT AlNMENT CLOSURE esta bli s hed. A ND AND b. RPV level le ss th an -126 in. (TAF). b. (R eactor vesse l/R CS [PWR] or RPV [BWR]) level l ess th a n CSl.3 a. RPV leve l cannot be monitored for 30 minutes or l o n ge r. (s i te-spec ifi c l eve l). AND (3) a. (R eactor vessel/RCS

[PWR] or RP V [BWR]) le vel cannot b. Core un covery i s indicated by EITHER of the fo llowin g: be monitored for 30 minute s or l o n ger.

Refueling Floor radiation mon it o r reading greater than AND R/hr b. Core un covery is indicated by ANY of th e fo ll ow in g:

UNPLANNED rise in drywe]l floor or equipment drai * (Sit e-s pecific radiation monitor) reading greater than sump le vels of sufficient magnitude to indicate core (si te-s pecific val ue) unco very

Erratic so urce ran ge monitor indication

[PW R]

UNPLANNED in crease in (site-specific s ump and/or tank) l evels of s ufficient magnitude to indicate core un covery * (Oth e r s ite-specific indications) Difference I Deviation I Justification 17 Difference:

COLD SHUTDOWN/

REFUELING SYSTEM MALFUNCTION ICS/EALS NEI 99-01 Rev 6 Thresholds (l)a and (2)a specify CONTAINMENT CLOSURE. MNGP Thresholds CSl.1.a and CSl.2.a specify SECONDARY CONTAINMENT.

Justification:

SECONDARY CONTAINMENT is equivalent to CONTAINMENT CLOSURE; therefore, the intent ofNEI 99-01Rev6 is still satisfied.

Difference:

NEI 99-01 Rev 6 Threshold (3)b, last bullet, refers to Other site-specific indications of core uncovery.

No Other site-specific indications are used at MNGP. Justification:

MNGP does not use any other site-specific indications of core uncovery other than those already identified in Threshold CSl.3.b. Difference:

Justification:

Site specific information provided.

Information is identified in MNGP EOP C.5.1-1100 and ARP C.6-065-A-06.

CA-04-202 provides the calculated Refueling Floor radiation monitor value. Plant design difference, -47 in. is used for EAL Threshold CSl.lb since MNGP does not have a Low ECCS actuation setpoint.

The intent of NEI 99-01 Rev 6 is still satisfied since this threshold is also linked with SECONDARY CONTAINMENT not being established.

18 COLD SHUTDOWN/

REFUELING SYSTEM MALFUNCTION ICS/EALS CAl: INITIATING C O NDITI O NS NEI 99-01 Rev 6 Montice ll o Nuclear Generating Plant Los s of (reactor vessel/RCS

[PW R] or RPV [BWR]) inventory. Lo ss of RPV inventory. Difference I Deviation I Justification None THRESHOLDS NEI 99-01 R ev 6 Montice ll o Nuclear Generating Plant (l) Lo ss of (reactor vesse l/RCS [PWR] or RPV [BWR]) in ve nt o r y as CAl.l Los s of RPV inve n tory as indi cated by level l ess t h an -47 in. indicated by l eve l le ss than (s ite-s pecific l evel). CAl.2 a. RPV level cannot be monitored for 15 minutes or longer (2) a. (Reactor vessel/RCS

[PWR] or RPV [BWR]) le vel cannot AN D be monjtored for 15 minute s or long er b. UNPLANNE D rise in drywell floor or equipment drain sump AND l eve l s due to a loss of RPV inventory. b. UN PL ANNE D increase in (site-s p ecific s ump a nd/or tank) le ve l s due to a lo ss of (reactor vessel/RCS

[PWR] or RPV [BWR]) in ve nt o r y. Difference I Deviation I Justification Difference:

Site specific information provided.

Justification:

Information is identified in MNGP ARP s C.6-003-A-38 and C.6-065-A-06.

19 COLD SHUTDOWN/

REFUELING SYSTEM MALFUNCTION ICS/EALS CA2: INITIATING CONDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Loss of all offsite and all onsite AC power to emergency buses for 15 minutes Loss of all offsite and all onsite AC power to essential b u ses for 15 minutes or l o n ger. or l o n ger. Difference I Deviation I Justification Difference:

NEI 99-01 Rev 6 IC refers to emergency buses. Monticello IC refers to essential buses. Justification:

Terminology difference

-the essential buses at Monticello are the emergency buses. THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (I) Loss of ALL offsite and ALL ons it e AC Power to (site-specific CA2.l Loss of ALL offsite and ALL onsite AC Power to essential buses 15 emergency buses) for 15 minutes or l onger. and 16 for 15 minutes or longer. Difference I Deviation I Justification Difference:

Site s pecific information provided.

Justification:

Ops Man B.09.06-01 identifies the essential buses. Terminology difference

-the essential buses at Monticello are the emergency buses. 20 COLD SHUTDOWN/

REFUELING SYSTEM MALFUNCT I O N I CS/E A LS CA3: INITIATING CONDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Inability to maintain the plant in cold shutdown.

Inability to maintain the plant in cold s hutdown. Difference I Deviation I Justification None THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant ( l) UNPLANNED increase in RCS temperature to greater than (site-CA3.I UNPLANNED ri s e in RCS temperature to greater than 12 °F for specific Technical Specification cold s hutdown temperature limit) grea ter than the duration s pecified in a ble C . for g reate r than the duratio n specified in the following table. Table C2: RCS Heat-up Duration Thresholds Table: RCS Heat-up Duration Thresholds RCS SEC O N D A R Y Heat-up Containment Closure Heat-up CONTA I NMENT Duration RCS Status Status Dur at ion Not Established 0 minutes Not intact Int ac t (but not at reduced Established 20 minutes* inventory

[PWR]) o t applicab l e 60 minutes* Inta ct N I A 60 minutes* No t intac t (or a t reduced Establ i shed 20 minute s*

lf an RCS heat removal system is in operation w i thin thi s time frame and invent ory [PWR]) Not Establ i s hed 0 minute s RCS temperature is bein g r ed uced , th e EAL i s not a ooli cab le. *If an RCS heat removal sys t e m is in opera tion wit hin this time frame and RCS temoerature i s being redu ce d , th e EAL i s not ao oli cab l e. CA3.2 UNPLANNED RCS pressure i se grea ter than 10 p s i cr. (2) UNPLAN ED RCS pressure increase greater th a n (site-specific pressure reading). (This EAL does not apply during water-solid pl ant conditions.

[PWR]) Difference I Deviation I Justification Difference:

Table d esignator C2 assigned to RCS Heat-up Duration Thresho l ds table. Threshold CA3.1 refers to Table C2. Justification:

Editorial change to clearly identify tables within the document.

Difference:

Information included in RCS Heat-up Duration Thresholds Table for Monticello is inverted from the pre se ntation in NEI 99-01 Rev 6. 21 COLD SHUTDOWN/

REFUELING SYSTEM MALFUNCTION ICS/EALS Difference:

Information included in RCS Heat-up Duration Thresholds Table for Monticello is inverted from the presentation in NEI 99-01 Rev 6. Information is the same. Justification:

Editorial change for Human Factors considerations -worst case is presented first. Difference:

The NEI 99-01 Rev 6 RCS Heat-up Duration Thresholds Table refers to Containment Closure. The MNGP RCS Heat-up Duration Thresholds Table refers to SECONDARY CONTAINMENT.

Justification:

Editorial change due to MNGP terminology-Containment Closure is equivalent to the SECONDARY CONTAINMENT at MNGP. Difference:

Site specific information provided.

Justification:

Cold Shutdown Temperature is listed in Technical Specification (TS) Table 1.1-1. Difference:

Site specific information provided for EAL threshold CA3.2. Justification:

The 10 psig value is based on NEI 99-01 Rev 6 Developer Notes guidance.

22 COLD SHUTDOWN/

REFUELING SYSTEM MALFUNCTION ICS/EALS CA6: INITIATING CONDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Hazardous event affecting a SAFETY SYSTEM needed for the current Hazardous event affecting a SAFETY SYSTEM needed for the cmTent operating mode. operating mode. Difference I Deviation I Justification None THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) a. The occurrence of ANY of the fo llowing hazardous events: CA6.l a. The occurrence of ANY of the following hazardous events:

Seismic event (earthquake)

Seismic event (Eart hquake)

Internal or external flooding event

Internal or external flooding event

High winds or tornado strike

Hi g h winds or tornado strike

FIRE

FIRE

EXPLOSION

EXPLOSION

( si le-specific hazard s)

River level greater than 919 ft el.

Other events with simi lar h azard characteristics as

River level less than 900.5 ft el. determined by the Shi f t Manager

Other events with simi lar hazard characteristics as AND determined by the Shift Manager b. EITHER of the following:

AND 1. Event damage has caused indications of degraded b. EITHER of the following:

performance in at lea st one train of a SAFETY

Event d a m age has caused in dication s of degraded SYSTEM needed for the current operating mode. perform a nce in at le as t one train of a SAFETY SYSTEM OR needed for the current operating mode. 2. The event ha s caused VISIBLE DAMAGE to a

Th e event has caused VISIBLE DAMAGE to a SAFETY SAFETY SYSTEM component or str ucture SYSTEM component or struct ure needed for the current needed for the current operating mode. operating mode. Difference I Deviation I Justification 23 COLD SHUTDOWN/

REFUELING SYSTEM MALFUNCTION ICS/EALS Difference:

MNGP Threshold CA6.lb. -deleted OR and replaced numbered indicators with bulleted listing. Justification:

Editorial revision -consistent use of the nesting within the EAL thresholds.

Intent of NEI 99-01 Rev 6 threshold remains satisfied.

Difference:

Site specific information provided.

Justification:

Seismic indication is identified in ARP C.6-006-C-08 and C.6-006-C-13.

MNGP procedure A.6 identifies the criteria for high and low river levels. 24 COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS CUl: INITIATING C O NDITI O NS NEI 99-01 Rev 6 Montice ll o Nuclear Generating Plant UNPLANNED lo ss of (reactor vessel/RCS

[PWR] or RPV [BWR]) inventory UNPLANNED lo ss of RPV inventory for 15 minutes or longer. for 15 minutes or lon ger. Difference I De v iation I Justification None THRESH O L D S NEI 99-01 Rev 6 Monticello Nuclear Generating P l ant (1) UNPLANNED loss of reactor coolant results in (reactor vesse l/RCS CUl.l U PLANNED lo ss of reactor coolant results in RPV level less than [PWR] or RPV [BWR]) level less than a required lower limit for 15 a procedurally required lower limit for 15 minutes or longer. minutes or l o n ger. CUl.2 a. RPV level cannot be monitored.

(2) a. (R eactor vessel/RCS [PWR] or RPV [BWR]) level ca nnot be AND monitored.

b. UNPLANNED rise in drywell floor or equipment drain sump AN D levels. b. UNPLANNED increase in (s it e-spec ific sump and/or tank) l eve l s. Difference I Deviation I Justification Difference:

Site s pecific information pro v ided. Justification:

ARP C.6-065-A-06 provides information to support EAL threshold CU1.2b. 25 COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION I C S/EALS CU2: INITIATING C O NDITIONS N EI 99-01 Rev 6 Montice ll o N ucl e ar Gener a tin g Pl a n t Loss of all but one AC power source to emergency buse s for 15 minutes or Lo ss of all but one AC power source to essential bu ses for 15 minute s or longer. longer. Difference I De v iation I Justification Dif fe r e nc e: NE I 99-0 1 R ev 6 I C r e f e r s to e m e r ge nc y bu ses. M ontic e llo I C r e f e r s t o esse nti a l bu s e s. Justification:

Terminology difference

-the essential buses at Monticello a r e the emergency buses. THRE S HO L D S N E I 9 9-01 Re v 6 M ontic e llo N ucl ear G e n e r a tin g Pl a n t ( l) a. AC power capability to (site-specific emergency buses) is CU2.l a. AC power capability to essentia l b u s e s 15 an d 16 is reduced to reduced to a single power source for 15 minutes or l onger. a single power source (T a ble S l) for 15 min u tes or longer. AN D A ND b. A n y addjtional sing l e power source failure wi ll res u lt i n b. Any additio n al s i ng l e power source failure wi ll result in loss of Joss of a ll AC power to SAFETY SYSTEMS. all AC power to SAFETY SYSTEMS. f[able Sl IR R eserve T ra n sfo r me r l AR R eserve Tra n sformer i,2 R Au xi li a r y Tra n sfo rm er # 1 1 Eme r gency D ie s el G enerator #12 E mergency D iesel Gepe rator Di ffe r e nc e I D ev iat i on I Justification Diff e rence: Sit e s pecific informatio n pro v ided. Justification:

Table Sl incorporates EPFAQ 2015-015 guidance.

Ops Man B.09.06-01 identifies the essential buses for MNGP. Terminology difference

-the essential buses at MNGP are the emergency buses. 26 COLD SHUTDOWN/ REFUELING SYSTEM MALFU N CTION I CS/EALS CU3: INITIATING C O N D ITI O NS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant UNPLANNE D increase in RCS temperature.

Unplanned rise in RCS temperature.

Difference I Deviation I Justification None THRESH O LDS NEI 99-01 Rev 6 Monticello Nuclear Ge n erating Plant (I) UNPLANNED increase in RCS temperature to greate r than (site-CU3.l UNPLANNED rise in RCS temperature to greater than 212 °F. specific Technical Specification cold shutdown temperature limit). CU3.2 L oss of ALL RCS temperature and RPV level indication for 15 (2) Loss of ALL RCS temperature and (reactor vessel/RCS

[PWR] or minutes or longer. RPV [BWR]) level indication for 15 minutes or longer. Difference I Deviat i on I Justification Difference:

Site specific information provided.

Justification:

The Cold Shutdown Temperature is listed in MNGP TS Table 1.1-1. 27 COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS CU4: INIT I ATING CONDITIONS NEI 99-01 Rev 6 Montice ll o Nuclear Generating Plant Lo ss of Vital DC power fo r 15 minutes or longer. Lo ss of Vital DC power for 15 minutes or l onger. Difference I Deviation I Justification None THRESHOLDS NEI 99-01 R ev 6 Montice ll o N uclear Generating Plant (l) Indi cated vo lt age i s l ess than (s it e-specific bus vo lt age va lu e) on CU4.l Indi cated vo lt age is le ss t h a n 110 VDC o n r equ ir e d 125 VDC Vit a l req uired Vital DC bu ses for 15 minute s or l o n ger. DC bu ses for 15 minute s or l o n ger. Difference I Dev i ation I Justification Difference:

MNGP Threshold CU4.1 identifie s specific Vital DC bu ses applicable to thi s threshold.

Justification:

The 125 VDC main distribution buses at Monticello are required to enable safe shutdown of the plant (USAR 08.05). Therefore, the intent of NEI 99-01 Rev 6 Threshold (1) is still satisfied.

Difference:

Site s pecific information provided. Justification:

The 110 VDC threshold is based on an average for both Division I and II batteries for battery voltages at 15 minutes prior to reaching the minimum required terminal voltage for equipment operation.

Calculations CA-02-179 and CA-02-192 provide the basis for the selected voltage. 28 COLD SHUTDOWN/

REFUELING SYSTEM MALF U NCT I O N I CS/EALS CVS: INITIATING CONDITIONS N E I 99-01 Re v 6 M ontic e llo N ucl ear G e n e r a tin g Pl a nt Lo ss of all onsite or offsite comm unication s capa bilitie s. L oss of all o n s ite o r offsite communications ca p abilities. Differen ce I De v iati o n I Justification Non e THRESHOLDS NEI 99-01 Re v 6 M onticello N ucle a r Gener a tin g Pl a nt ( 1) Loss of AL L of the fo ll owing onsite co mmunic a ti on m e thods: CUS.l Lo ss of AL L of the fo ll ow ing o n s it e co mmunic at ion m et hod s: (s ite-s pecific l i s t of communications method s)

Commercial Telephones (2) Loss of A LL of the following O R O comm u ni ca tion s method s:

Plant Telephones (s ite-specific l i s t of communications method s)

Portable radios (3) Loss of ALL of t h e following N R C communications method s:

Plant PA Syste (site-specific l i s t of communications method s) CUS.2 Lo ss of AL L of the fo ll owing Offsite Re s pon se Orga n i za tion (ORO) communications m e t h ods:

Commercial T elep h ones

Direct Dedicated Telephones

Radio/Receiver Transmitter CUS.3 Lo ss of A LL of the following NRC comm u nic at io n s method s:

Federal Telecommunications System (FIS)

Commercial Telephones Difference I De v iation I Justification Diff e r e nce: Site s p e cific information pro v id e d. 29 INDEPENDENT SPENT FUEL STORAGE FACILITY (ISFS I) I CS/EALS EUl: INIT I ATING C O NDITI O NS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Damage to a loaded cask CONFINEMENT BOUNDARY. Dam age to a loaded cask CONFINEMENT BOUNDARY. Difference I De v iation I Justification None THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (I) D amage to a l oa d ed cask CONFINEMENT BOUNDARY as EUJ.l D amage to a l oa d ed cask CONFINEMENT BOUNDARY as indicated b y an on-contact rad i at ion reading greater th a n (2 time s the indicated b y an on-contact radiat i on readi n g greater than any o f th e s ite-s pecific cas k s pecific technical specification a llow a ble radiation values li sted in Table El: l eve l) on the s urfa ce of the s p e nt fuel cask. Table El [ocation of Dose Rate _Jf otal Dose Rate (Neutron+

Gamma mR/hr) HSM or HSM-H Front Surface 1400 HSM or HSM-H Door QOO End shield wall exterior 40 Difference I Deviation I Justification Difference:

Added new Table El to MNGP Threshold EUl.1. Site s pecific information provided at twice the ISFSI TS 1.2.7f va lue. Justification:

Utilized table to display ISFSI technical specification radiation levels for the different ISFSI modules. Intent of NEI 99-01 Rev 6 EAL threshold remains satisfied.

30 FISSION PRODUCT BARRIER ICS/EALS . : . o';... *' .* . __ ,-; BWR FISSION PRODUCT .BARRIER MATRIX.; INITIATING CONDITIONS/THRESHOLDS

"-* * ---> ' " ' --' *\' -,',,,,,";_,'

.. * .. NEI 99-01 Rev 6 F Al -Any loss or any Potential Loss of either the Fuel FSl -Loss or Potential Loss of any two barriers.

FGl -Loss of any two barriers and Loss or Potential Clad or RCS barrier Loss of the third barrier. * .. .. . Mopticello Nuclear*Generat_ing Plant .. FGl -Loss of any two barriers and Loss or Potential FSl -Loss or Potential Loss of any two barriers.

FAl -Any Loss or any Potential Loss of either the Loss of the third barrier. Fuel Clad or RCS barrier. Difference I Deviation I Justification

. None *. *. Fuel Cfad Barrier RCS Barrier Containment Barrier Loss Potential Loss Loss Potential Loss Loss Potential Loss NEI 99-01 Rev 6 1. RCS Activity 1. Primary Containment Pressure 1. Primary *containment Conditions A. (Site-specific Not Applicable A. Primary containment Not Applicable A. UNPLANNED rapid A. Primary indications that pressure greater than drop in primary containment reactor coolant (site-specific value) containment pressure pressure greater activity is greater due to RCS leakage. following primary than (site-specific than 300 µCi/gm containment pressure value) dose equivalent I-rise OR 131). OR B. (site-specific B. Primary containment explosive mixture) pressure response not exists inside consistent with primary LOCA conditions.

containment OR c. HCTL exceeded.

31 FISSION PRODUCT BARRIER ICS/EALS Monticello Nuclear Gene r ating Plant A. Cool a nt activity is Not App l ica bl e A. P ri m ary co nt ai nm en t No t App l icab l e A. UNPLANNE D r apid A. P rimary co nt a inm e n t great e r th a n 300 press u re greater than drop in primary pressure greater than µC i/gm do se 1.8 4 p s i g due to R CS containment pressure 56 p s i g equi val ent 1-13 1. leakage. following primary OR containme n t pressure B. Greater th a n or equal rise to 6% hydrogen and OR gre a ter th a n or equ a l B. Primary containment t o 5% oxyg en in pressure respo n se not IDr yw ell o r Toru s cons i stent w ith OR LOCA conditio n s. c. HC lJ exceede d. Diff e ren ce I De v iation I Justification Differenc e: S ite s pecific information pro v ided. Justification:

MNGP relies on radiochemistry analysis to determine dose equivalent 1-131 concentration as allowed by NEI 99-01Rev6.

MNGPEOP C.5.1-1200 provides guidance in response to primary containment pressure of 1.84 psig. C.5.1-1200 also identifies potential explosive mixtures of H 2 and 0 2 in the Drywell or Torus. The design pressure for the primary containment system at MNGP is 56 psig (USAR 5.2.1.1).

Terminology difference

-HCL is equivalent to HCTL at MNGP. NEI 99-01 Re v 6 2. RPV Water Le v el 2. RP V Wat er L ev el 2. RP V W at er L eve l A. P rimary containmen t A. RP V water level A. RPV water l evel Not App li cable Not Applicab l e A. Primary conta i nme n t flooding req u ired. can n ot be restored cannot be restored flooding r equired. a n d maintaine d above and maintained above (si t e-specific RP V (site-specific RP V water l eve l wa t er l eve l co r res p o nd i n g t o th e correspo n di n g t o th e to p of active fue l) or top of active fue l) or ca nn o t be d eterm in e d. ca nn ot b e determi n ed. Monticello N u clear Generating Plant 32 A. SAMG entry is required. A. RPV w a ter level cannot be re s t o red and maintain e d a b o v e -1 26in.or c ann ot b e d e t e rmin e d. Difference I Deviation I Justification FISSION PRODUCT BARRIER ICS/EALS A. RPV wa t e r level No t A ppli ca ble N ot A ppl i c a ble c a nnot b e r es tored and m ai ntained abo ve -126 in. o r c ann ot b e de t erm in e d. D i fference:

Fue l C l ad Barrier Loss EAL Threshold 2.A -added "SAMG entry is required".

A. Justification:

Revised EAL threshold based on EPFAQ 2015-004 guidance.

Terminology difference

-MNGP refers to the SAG as SAMG. Difference:

Containment Barr i er Potential Loss EAL Threshold 2.A -added "SAMG entry is required".

Justification

Revised EAL threshold based on EPFAQ 2015-004 guidance.

Terminology difference

-MNGP refers to the SAG as SAMG. Difference

Site specific information provided for RPV water level. Justification:

MNGP EOP C.5.1-1100 identifies the applicable RPV water level (TAF) for this threshold.

NEI 99-01 Rev 6 SAMG entry is regui r ed. 3. Not Applicable

3. RCS Leak Rate 3. Primary Containment Isolation Failure o t A ppli ca bl e N o t A ppli ca bl e A UNI SO L ABLE bre ak A. UN ISOLABL E A. U NI S OLABLE dir ec t No t A ppli ca bl e in ANY o f the prim ary sys t em d ow n s tr e am pa t hwa y fo ll ow in g: (s ite-l eakage t h at r es u lts t o the e n vir onm e n t s p ec i fic sys tems w ith in excee din g ex i s t s afte r prim a r y p o t e nti al fo r hi g h-EITHER o f th e co nt a inm e nt i so l a ti o n e n e r gy lin e bre aks) fo ll ow in g: s i g n a l OR l. Max No r ma l OR B. E m erge n cy RP V Op era tin g B. I n t en ti o n a l p ri mary De p r essuriza ti o n. T e mp era tu re conta i nme n t ve ntin g OR p e r E OP s 2. Ma x N o rm a l OR Ope ra tin g Area c. UNISOLABLE R a di a ti o n Le v el. primar y syste m l eakage th at r es ult s in excee din g EITHER o f the fo llo wi n g: 33 FISSION PRODUCT BARRIER ICS/EALS l. Max S afe Operati n g T e mperatur e. OR 2. Max Safe Operatin g Area R a diation Level. Montice ll o Nuclear Generating P l ant Not Applicable Not Applicable A. UNISOLABLE br eak A. UNISOLABLE A. UNISOLABLE dir ect Not Applicable in ANY of the prim ary syste m do w n strea m pathw ay following:

MSL; leakage that results to the environment HPCI; RWCU; RCIC in exceeding contro exists after prim a r y as indicated by hig room indications o f co ntainment iso lati o n ow/temperature EITHER of the s i g n a l i so l a tion s etpoints following:

OR OR I. Max Normal B. Intentional primary B. Emergency RPV Op erat in g containment ve ntin g Depr ess urization.

Temperature perEOPs OR OR 2. Max Normal c. UNISOLABLE Operatin g Area primar y syste m Radi atio n Level. l eakage that results in excee ding contro room indications 0£ EITHER of the fo llowing: l. Max S afe Operating Temperature.

OR 2. Max Safe Op era tin g Area R a diation Level. 34 FISSION PRODUCT BARRIER ICS/EALS Diffe r ence I Dev i ation I Justification Differe n ce: Si t e s p ecific informat i on provided for RCS Barr i er Loss 3A. Justification:

MNGP indication s for the loss of the RCS Barrier are included.

Difference:

Added 'contro l room indications of' RCS Barrier Potential Loss 3A and Containment Barrier Loss 3C. Justification:

Editorial change to specify those indications that are readily available to the decision maker to enable timely assessment and classification (EPFAQ 2015-012).

NEI 99-01 Rev 6 4. Primary Co n tainment R adiat i on 4. Primary Containme n t Rad i ation 4. Primary Containment Radiation A. Prim ary co nt ainme nt Not Applicable A. Prim ary co ntainm ent Not Applicable Not Applic a ble A. Prim ary co nt a inm ent radiation monitor radiation monjtor radiation monjtor readin g greater than reading g reater than reading greater than (site-specific value). (s ite-specific value). (s ite-s pecific va lue). Montice ll o N u clear Ge n erating P l ant A. Containment High Not Applicable A. Containment Hi gh Not Applicable Not Applicab le A. Containment High Range R ad (Drywe l ange Rad (Drywe ll Range Rad (Drywel adiat i on) monjtor ad i at i on) monitor adiation) monitor reading greater than reading greater than reading greater than 1.5 E+03 R/hr 6.2 E+O l R/hr 3.3 E+04 R/hr Diffe r ence I Dev i ation I Justification D i ffe r ence: MNGP Loss of Fuel Clad, Loss of RCS Barrier , and Potential Loss of Containment Barrier Thresholds 4.A specify Containment High Range Rad monitors.

Justification:

Plant design -Primary containment radiation monitors referenced in NEI 99-01 Rev 6 are the Containment High Range Rad monitors at MNGP. Differe n ce: Site s p ecific informa t ion p rovided. Justification:

Radiation Monitor Calculation CA-04-194 identifies the values for the Containment High Range Rad monitors.

35

5. Other Indications A. (site-specific as applicable).

A. (site-specific as applicable)

Difference I Deviation I Justification FISSION PRODUCT BARRIER ICS/EALS NEI 99-01 Rev 6 5. Other lndieations A. (site-specific as applicable)

A. (site-specific as applicable)

Generating giant . *'"' ,. "' ' " ,., ' -""" Other Imlieatians 5; Other Indications A. (site-specific as applicable)

Other fudieatiees A. (site-specific as applicable)

Difference:

No other indications that indicate a loss or potential loss of the fission product barriers have been determined.

Therefore this section has been deleted for MNGP. NEI 99-01 Rev 6 6. Emergency Director Judgment 6. Emergency Director Judgment A. ANY condition in the A. ANY condition in the A. ANY condition in the A. ANY condition in the opinion of the opinion of the opinion of the opinion of the Emergency Director Emergency Director Emergency Director Emergency Director that indicates Loss of that indicates that indicates Loss of that indicates the Fuel Clad Barrier. Potential Loss of the the RCS Barrier. Potential Loss of the Fuel Clad Barrier. RCS Barrier. . . :5.**

Judgment

' '" ' . '"" ' ,

A. ANY condition in the A. ANY condition in the A. ANY condition in the A. ANY condition in the opinion of the opinion of the opinion of the opinion of the Emergency Director Emergency Director Emergency Director Emergency Director that indicates Loss of that indicates that indicates Loss of that indicates the Fuel Clad Barrier. Potential Loss of the the RCS Barrier. Potential Loss of the Fuel Clad Barrier. RCS Barrier. 36 6. Emergency Director Judgment A. A. ANY condition in the opinion of the Emergency Director that indicates Loss of the Containment Barrier. A. ANY condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier.

i udglllent . , , ,, ]': . , . " ANY condition in the A. ANY condition in the opinion of the opinion of the Emergency Director Emergency Director that indicates Loss of that indicates the Containment Potential Loss of the Barrier. Containment Barrier.

FISSION PRODUCT BARRIER ICS/EALS Difference

/ Deviation/.Justificatil>n None 37 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SA F E TY I CS/EALS HGl: INITIATING C O NDITIONS NEI 99-01 Re v 6 Montice ll o Nuclear Generating Plant HOSTILE ACTION resulting in loss of physical control of t h e facility.

HOSTILE ACTIO res ultin g in loss of physical control of the facility.

Difference I Deviation I Justification None THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) a. A HOSTILE ACT ION i s occurr in g or ha s occ urr e d wi thin HGl.l a. A HOSTILE ACT I ON i s occurring or h as occurre d withj n the the PROTECTED AREA as repo rted by the (site-sp ecific !Plan PROTECT E D AREA as reported by the Security Shi sec u ri t y shift s u pervision). Supervisor.

AND AND b. EITHER of the following h as occ u rred: b. EITHER of th e fo ll owi n g has occurred:

l. ANY of the fo llowin g safe t y functions cannot be 1. ANY of the following safety functions cannot be co n tro ll ed or mruntained.

contro ll ed or maintained.

R eactivity contro l

R eactivity contro l

Core cool in g [PWR] I RP V water l evel [BWR]

RPV water le vel

RCS hea t remo val

RCS heat removal OR OR 2. D a mage to spe nt fuel h as occurred or i s IMMINENT.

2. D amage to s pent fuel ha s occ u rred o r i s IMMINENT.

Difference I Deviation I Justification D i fference:

Added clarifier "Plant" before PROTECTED AREA to MNGP Threshold HG I.la. Justification:

Human Factors consideration to clarify that this threshold does not apply to the ISFSI Protected Area which is applicable to MNGP Threshold HAl.1. Differ e nc e: Site s pecific information provided.

Justification:

Site-specific security shift supervision at MNGP is the Security Shift Supervisor.

38 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HG7: INITIATING CONDITIONS

.. </. ; . .. . NEl 99-01 Rev 6 Monticello Nuclear Generating Plant Other conditions exist which in the judgment of the Emergency Director Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency.

warrant declaration of a General Emergency.

Difference I Deviation I Justification None TllRESHOLDS NEl99*01 Rev.6 Monticello Nuclear Generating Plant (1) Other conditions exist which in the judgment of the Emergency HG7.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the ACTION that results in an actual loss of physical control of the facility.

Releases can be reasonably expected to exceed EPA facility.

Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the Protective Action Guideline exposure levels offsite for more than immediate site area. the immediate site area. Difference

/Deviation I Justification

.. None 39 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HSl: I NITIATING CONDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant HOSTILE ACTION within the PROTECTED AREA. HOSTILE ACTIO within the Pl an PROTECTED AREA. Difference I Deviation I Justification Difference:

Added clarifier "Plant" before PROTECTED AREA to Monticello IC. Justification:

Human Factors consideration to clarify that this IC does not apply to the ISFSI Protected Area which is applicable to Monticello IC HAL THRESH O L D S NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) A HOSTILE ACTION is occurring or ha s occurred w it hin the HSl.1 A HOSTILE ACTION i s occurring or h as occurred within the Plant PROTECTED AREA as reported by the (s ite-specific sec urity sh i ft PROTECTED AREA a reported by the Security Shift Supervisor.

supervision).

Difference I De viat ion I Justification Difference:

Added clarifier

" Plant" before PROTECTED AREA to MNGP Threshold HSl.1. Justification:

Human Factors consideration to clarify that this threshold does not apply to the ISFSI Protected Area which is applicable to MNGP Threshold HAl.1. Difference

Site specific information provided.

Justification:

Site-specific security shift supervision at MNGP is the Security Shift Supervisor.

40 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY I C S/EALS HS6: INITIATING C O NDITI O NS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Inability to control a key safety function from outside the Control R oom. Inability to control a key safety function from outside the Control R oom. Difference I Deviation I Justification Non e THRESHOLDS NEI 99-01 Rev 6 Montice ll o Nuclear Generating P l ant ( 1) a. A n eve nt h as res ul te d in pl a n t control be i ng tra n sferre d H S6.l a. An eve n t h as res ul ted i n pla n t co n tro l be in g tra n sfe r red from from t h e Co n trol R oom to (si t e-spec i fic re m ote sh u tdow n t h e Control R oom to the alternate shutdown pane . pa n e l s and l oca l co ntr o l sta ti ons). AND AND b. Co n tro l o f ANY of t h e fo ll ow in g key safe t y fun c ti o n s is n ot b. C o n tro l of ANY of th e fo ll ow in g key safe ty f u nct i o n s i s n o t r eesta b lis h e d w ithin I 0 minutes. reesta bli s h e d w i t hin (s it e-s p ecific numb e r of mi nu tes).

R eac ti vity contro l (Modes 1 and 2 only)

R eact i v it y co n tr ol

RP V wa t er l evel

Core coo l i n g [P W R] I RP V wa t er l eve l [BW R]

R CS h ea t removal

RCS h ea t re m ova l Difference I De v iation I Justification Difference:

Site s pecific information provided.

Justification:

Abnormal Procedure (AOP) C.4-C identifies requirements for the establishment of control at the alternate shutdown panel. Deviation: Reactivit y Control identified in fir s t bulleted item of NEI 99-01 Re v 6 EAL Threshold (l)b i s applicable in A ll mode s. Reacti v it y Control identified in fir s t bulleted item of MNGP EAL Threshold HS6.lb i s applicable to Mode s 1 and 2 onl y. Justification:

Implements EPFAQ 2015-014 guidance.

Implementation of this guidance has been determined to be an acceptable deviation by the NRC (ML16166A240).

41 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS . ' HS7: INITIATING

., '" . . .. . . .. NEI 99-01 Rev 6 Monticello Nuclear Generating Plant .*.* Other conditions exist which in the judgment of the Emergency Director Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency.

warrant declaration of a Site Area Emergency . ...... Differen<:e I Deviation I Justification None THRESHOLDS

- . NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) Other conditions exist which in the judgment of the Emergency HS7.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. effective access to equipment needed for the protection of the Any releases are not expected to result in exposure levels which public. Any releases are not expected to result in exposure levels exceed EPA Protective Action Guideline exposure levels beyond the which exceed EPA Protective Action Guideline exposure levels site boundary.

beyond the site boundary.

Difference I Deviation I Justification None 42 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HAl: INITIATING CONDITI O NS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes. attack threat wit hin 30 minutes. Difference I Deviation I Justification None THRESH O LDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) A HOSTILE ACTION is occ urrin g or has occ urr ed within the HAl.l A HOSTILE ACTION is occurring or has occmTed within the OWNER CONTROLLED AREA as reported by the (site-specific OWNER CONTROLLED AREA as reported by the Security Shift sec urity s hift s upervision). Su e rvisor. (2) A validated notifi cat i o n from NRC of an a ir craft attack threat w ithin HAL.2 A val.idated notification from NRC of a n aircraft attack threat within 30 minutes of the si t e. 30 minutes of the site. Difference I Deviation I Justification Difference

Site specific information provided.

Justification:

Site-specific security shift supervision at MNGP is the Security Shift Supervisor.

43 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HAS: INITIATING CONDITIONS N E I 99-01 Rev 6 Monticello Nucle a r Generatin g Pl a nt Gaseo u s release impeding access to equipment neces sar y for norma l p l ant G aseo u s release impeding access to equipment nece ssary for normal p l ant opera t ions , cool down or sh u tdown. opera t ions, coo l down or s hu tdown. Difference I Deviation I Justification Non e THRESHOLDS N E I 99-01 Rev 6 Monticello N ucle a r Generatin g Plant (1) a. R elease of a toxic, corrosive , asp h yx i a n t or flammab l e gas HAS.I a. R elease of a t oxic , corrosive, asp h yxiant or flammab l e gas into i nto any of t h e fo ll owing p l a n t rooms or areas: a n y of the Table ITT p l a nt rooms or areas: (site-specific l is t of p l ant rooms or areas with e n try-re l ated Table H!l mode app li cab i lity identified)

Building Rooms Applicable A ND Mode(s) b. Ent r y i n t o the room or area i s pro hi bi t ed or imp e ded. R eaoto r Buildin g All All Turbine Building All All Int a ke Structure All All A ND b. Entry i n t o th e room or area is prohib i ted or i mp eded. Diff e r e nc e I De v iati o n I Justification Diff e r e nc e: NE I 9 9-01 R ev 6 Thr es hold (l)a do es n o t includ e tabl e r e f ere n ce. MNG P T h res h o ld H A S.l a includ es r e f e r e n ce t o T a bl e Hl a nd in c orp o r a t es th e tabl e. Justification:

Human Factors consideration

-use of table format clearly identifies the applicable rooms and plant mode s. Diff e r e nc e: Site s p e cific information pro v id e d. Justification:

The areas listed in Table Hl have been determined by MNGP personnel as requiring access for essential operations that must be performed during normal operations and to reach cold shutdown.

4 4 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HA6: INITIATING CONDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Co n trol R oo m evac u a ti on r es ulting i n tr a n sfe r of pl a nt co n tro l to a l te rn a t e Control R oo m evac u a ti on r es ult ing i n tr a n sfe r of pl a nt co n tro l to a lt e rn a t e l oca ti o n s. l oca ti o n s. Difference I Deviation I Justification None THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) An eve nt h as res ul te d i n pl an t co ntr o l be in g transfe rr e d from t h e HA6.1 A n eve nt h as res ul te d in !a nt co ntr ol b e in g tra n sfe rr e d fro m th e Control Ro o m to (s it e-s p ec ifi c rem o te s hutd o wn p a nel s a nd l oca l Control R o om to the alternate s hutdown p a n e l. co n tro l s t a ti o n s). Difference I Deviation I Justification Difference:

Site s pecific information provided. Justification

The alternate shutdown panel is the site-specific remote shutdown panel for MNGP. AOP C.4-C provides additional information related to operations from this location.

45 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert. Difference I Deviation

/Justification None (1) Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Differe11ce I Deviation

/Justification None Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert. HA 7 .1 Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. 46 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HUl: INITIATING C O NDITIONS NEI 99-01 Rev 6 Montice ll o Nuclear Generating P l ant Confirmed SECURITY CONDITION or threat. Confirmed SECURITY CONDITION or threat. Difference I Deviation I Justification None THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating P l ant (1) A SECURITY CONDITION th a t doe s not involve a HOSTILE HUl.l A SECURITY CONDITION that doe s not in vo l ve a HOS TILE ACTION as reported by the (si t e-specific s ecurity s hift s upervi s ion). ACTION as reported by the Se c urity Shi ft t Su ervisor. (2) otification of a credib l e sec urit y threat directed at the s ite. HUl.2 Notification of a credible sec urity threat directed a t MNGP. (3) A validated notification from the NRC providing information of an HUL.3 A va]jdated notification from the NRC providing information of a n ai r craft thre a t. aircraft threat. Difference I Deviation I Justification Difference:

MNGP Thresho l d HUl.2; replaced 'the site' with MNGP. Justification:

Editorial change -clearly identifies that threat is directed against the Monticello site. Difference:

Site specific information provided.

Justification:

Site-specific security shift supervision at MNGP is the Security Shift Supervisor.

47 HAZARDS AND OTHER CONDITIONS AFFECTING PLAN T SAFETY ICS/EALS HU2: INITIATING CONDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Se i smic eve nt g r ea t e r th an OBE l eve l s. S eismic eve nt g r ea t e r th a n OB E l eve l s. Difference I Deviation I J u stification None THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) Se i s mi c eve nt grea t e r th a n Op era tin g B asis Ear thqu ake (QB E) as H U2.I S e i s mi c eve nt grea t e r th a n Op e ratin g B as i s Ear thqu ake (QB E) as indi ca ted b y: indi ca t e d b y Annun o iator OPERA11IONAL BASIS EARTHQUAKE (si t e-s pecifi c indic a ti o n th a t a se i s m ic eve nt m e t o r excee ded OBE (6-C-13) r ece ived. limit s) Difference I Deviation I Justification Difference:

Site specific information provided.

Justification:

Seismic events are indicated in the Control Room by annunciator 6-C-13. ARP C-6-006-C-13 provides guidance on response to seismic events. 48 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HU3: INITIATING CONDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant H aza rd o u s e ve nt. H aza rd o u s eve nt. D i fference I Deviation I Justification Non e THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (I) A t orna d o s tr ike w i t hin th e PRO TECTE D AREA. H U3.l A tornado st r ike w i t hin t h e Plant PRO TECTE D AREA. (2) Int e rn a l roo m o r area fl oo din g of a m ag nitude s u fficient to re qu ire H U3.2 Int e rn a l roo m o r area floo din g of a m ag n_i t ude s u ffic i e n t to re qu ire m a nual or a utomatic e l e ctri cal i so l a ti o n o f a S AF ETY SYSTEM m a nu a l or a utom a tic e le c tri ca l i s ol a ti o n of a S AFE TY SYST E M c o mponent n ee ded fo r the c urr e nt o p era tin g m o d e. co mp o n e nt n ee d e d fo r th e c un-e nt o p era tin g m o d e. (3) M ove m e nt of pe r so nn e l wit hin t h e PROT E CT E D AR EA i s imp ede d H U3.3 Move m e nt of p e r so nn e l w i t hi n the Pl a n t PRO TE C TE D A RE A i s du e t o an offs it e eve nt in vo l v in g h azar d o u s ma t e ri a l s (e.g., a n offs it e imp ede d du e t o an offs it e eve nt in vo l v in g h azar d o u s mater i a l s (e.g., c h e mi ca l s pill or t ox i c gas re l ease). a n offs ite c h e mi ca l sp ill or t ox i c gas r e l ease). (4) A h azar d o u s eve nt th at r es ults i n o n-s i te co nditi o ns s u ffic i e n t t o H U3.4 A h azar d o u s eve nt th at r es ul ts in o n-s it e co n d iti o n s s u fficie nt t o prohibit th e pl a nt s t aff fro m access in g the s it e vi a p e r so n a l v e hi c l es. p ro hibit th e pl a nt s t aff fr om access in g the s it e vi a p e r so n a l ve hicl es. (5) (Sit e-specific li s t of n a tur a l or tec hn o l og i ca l h azar d eve n ts) HU3.5 Ri ver l eve l g r eater t h a n 9 18 f t e l. HU3.6 River L eve l I es-s th a n 902.4 ft e l. Difference I Deviation I Justification Difference:

Added clarifier

" Plant" before PROTECTED AREA to MNGP Threshold HU3.1. Justification:

Human Factors consideration to clarify that this threshold does not apply to the ISFSI Protected Area. Difference:

Site specific information provided.

Justification:

MNGP procedure A.6 provides the criteria for river levels identified in EAL thresholds HU3.5 and HU3.6. 49 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HU4: INITIATING C O NDITIONS NEI 99-01 Rev 6 Monticello Nuclear Gen erat ing P l ant FIRE potentially d egra din g the l eve l of safety of the plant. FIRE potentially degrading th e l eve l of safe ty of the plant. Difference I Deviation I Justification None THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant ( l) a. A FIRE i s NOT ext in g ui s h e d within 15-minut es of ANY of HU4.l a. A FIRE i s NOT ext in g ui s h e d within 15-minut es of ANY of the the following FIRE detecti o n indic a ti o n s: following FIRE d etect ion indications:

R e port from the field (i.e., v i s u a l observation)

Report from the field (i.e., v i s u a l observation)

R eceip t of multiple (mo r e than l) fire a l ar m s or

Receipt of multiple (m o r e than l) fire alarms or indication s indi ca tion s

Fi e ld verification of a s in g l e fire a l ar m

Field verification of a s in g le fire alarm AND b. The FIRE i s lo ca ted within ANY of the Table H2 pl a nt rooms or AND areas. b. The FIRE i s located within ANY of the fo ll ow in g plant rooms or a r eas: HU4.2 a. Receipt of a s in g le fire alarm (i.e., n o other indicati o n s of a FIRE). (site-specific li s t of plant rooms or area s) AND (2) a. R ece ipt of a s in g l e fire alarm (i.e., no other indications of a b. The FIRE is locat e d within ANY of the Table H 2 pl a nt rooms or FIRE). areas. AND AND b. The FIRE i s loc ated wit hin ANY of the following plant c. The existence of a FIRE i s not ve rifi e d wi thin 30-minutes of a larm room s o r areas: receipt. (si t e-spec ifi c li st of plant rooms or areas) AND HU4.3 A FIRE within the Plant PROTEC i I1E D AREA or ISFSI PROTECTED

c. The existence of a FIRE is not verified within 30-minutes of AREA not extinguished within 60-minut es of the initi a l report, a l arm or alarm receipt. indic a tion. (3) A FIRE wit hin th e plant or ISFSI [for p lants w ith an I SFSI outside the plant Prot ecte d Ar ea] PROTECTED AREA n ot ex tin g ui s hed within HU4.4 A FIRE within t h e Plant PR0 1D EC11ED AREA or ISFSI PROTECTED 60-minutes of th e initi a l report , alarm or indic a tion. AREA that require s firefighting s upport by an offsite fir e r espo n se (4) A FIRE within the pl a nt or IS FSI [for plants with an ISFSI outside th e agency to extinguish.

50 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS plant Prot e cted Ar e a] PROTECTED AREA that require s firefighting support by an offsite fire response agency to extinguish. Table H2 Building Name Room(s)/Area(s) with Safe t y Equipment Reactor Buildin g All HPCI Building All Turbine Building All Control a nd Control Room , Cable Administration Buildin g s r eacl in g Room , and B a tt ery Rooms Diesel Generator All Build i ng Die sel Fuel Oil 1fr.a n sfeu All House EFT Building All Intak e Structure A!l l Difference I Deviation I Justification Differences:

MNGP Thresholds HU4.lb and HU4.2b -added reference to Table H2 instead of listing areas separately for each threshold.

Justification:

Human factors consideration

-applicable rooms are the same for each threshold.

Placing these rooms into one table and referencing that table in the threshold simplifies the process for identifying applicable rooms. Difference:

MNGP Thresholds HU4.3 and HU4.4 include PROTECTED AREA after Plant. Justification:

Editorial change for clarity. Intent of NEI 99-01Rev6 EAL threshold (3) and (4) is still satisfied.

Differences:

Site specific information provided.

Justification:

Human factors consideration

-Table H2 identifies rooms/areas that contain safety systems. 51 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS Other conditions exist which in the judgment of the Emergency Director warrant declaration of a (NO)UE. Differeµce I Deviation/

Justification*

None (1) Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs. *Difference I Devfation I Justification None Other conditions exist which in the judgment of the Emergency Director warrant declaration of a NUB. HU7.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring off site response or monitoring are expected unless further degradation of safety systems occurs. 52 SYSTEM MALFUNCTIONS SGl: INITIATING C O NDITI O NS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Prolonged l oss of all offsite a nd a ll onsite AC power to e m erge ncy bu ses. Prolon ge d l oss of a ll offsite a nd a ll o n si te AC power t o essential bu ses. Difference I Deviation I Justification Difference

NEI 99-01 Rev 6 IC refers to emergency buses. MNGP IC refers to essential buses. Justification:

Terminology difference

-the essential buses at MNGP are the emergency buses. THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) a. L oss of ALL offs it e an d ALL onsi t e AC p owe r to (s ite-SGl.l a. Lo ss of ALL offsite a nd ALL o n s ite AC power to e s se nti a l s p ec ific emerge n cy bu ses). buses 15 and 16. AND AND b. EITHER of th e fo ll owing: b. EITHER of th e fo ll owing:

R es to ra ti o n of at le as t o n e AC e m e r ge n cy bu s in l ess

R es t oration of a t l eas t one AC esse nti a l bus i n l ess th a n th a n (s it e-s pe c i fic hour s) is n o t li ke l y. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months i s n o t likel y. * (Site-specific indicati o n of an in a bility t o adeq u a t e l y

Rea o tor ve sse l water lev e l ca nnot be r es t o r e d an d remove he a t fro m the core) m a int a ined a bo ve -1 49" (Minimum Steam Cooling RPV Water L eve l) Difference I Deviation I Justification Difference:

Site s pecific information provided. Justification:

Ops Man B.09.06-01 identifies the essential buses for MNGP. EOP C.5.1-1100, Part J, provides the basis for Minimum Steam Cooling RPV Water Level (-149"). Terminology difference

-the essential buses at MNGP are the emergency buses. MNGP USAR-08.12 identifies MNGP as a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months coping plant. 53 SYSTEM MALFUNCTIONS SGS: INITIATING CON D ITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Loss of a ll AC and Vital DC power so u rces for 15 minutes or longer. Loss of all AC and Vital DC pow e r so urces for 15 rninutes or longer. Difference I Deviation I Justification None TH R ESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant ( l) a. Lo ss of ALL offsite and ALL onsite AC power to (s ite-SG8.1 a. Lo ss of ALL offs it e a nd ALL onsite AC power to e s-se nti a l specific emergency buses) for 15 minutes or l o n ger. bu ses 15 and 16 fo r 15 minute s or l o n ge r. AND AND b. Indicated voltage i s le ss th a n (site-s pecific bu s vo lt age b. Indicated voltage i s l ess th a n 110 VOC on ALL 1 2 5 VOC va lu e) on ALL (site-specific Vital DC bu ses) fo r 15 Vital DC bu ses for 1 5 minutes or l onger. minutes or l onger. Difference I Deviat i o n I Justification Difference:

Site s pecific information provided.

Justification:

Ops Man B.09.06-01 identifies the essential buses for MNGP. Essential buses 15 and 16 are the MNGP emergency buses. The 110 VDC threshold is based on an average for both Division I and II batteries for battery voltages at 15 minutes prior to reaching the minimum required terminal voltage for equipment operation.

Calculations CA-02-179 and CA-02-192 provide the basis for the selected voltage. USAR-08.05 supports the use of 125 VDC Vital DC buses for this threshold.

54 SYSTEM MALFUNCTIONS SSl: INITIATING CONDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Loss of a ll offs ite a nd a ll o n s ite AC p ower to e m e r ge n cy b u ses fo r 15 minut es Loss of all offs i te a nd a ll o n si t e AC power t o esse ntial bu ses for 15 minut es o r l o n ge r. or l o n ge r. Difference I Deviation I Justification Difference:

NEI 99-0 1 R ev 6 IC r efe r s to e m e r ge n cy b u ses. MNGP IC r efe r s t o esse nti a l bu ses. Justification

Terminology difference

-the essential buses at MNGP are the emergency buses. THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant ( 1) L oss of ALL offsite a nd ALL o n s ite AC p ower t o (s i te-s p ec ifi c S S l.l L oss of ALL offs it e a nd ALL o n s i te AC p owe r to essential bu ses 15 e m e r g en cy bu ses) fo r 15 minut es o r l o n ge r. and 16 fo r 15 minut es or l o n ge r. Difference I Deviation I Justification Difference:

Site specific information provided. Justification:

Ops Man B.09.06-01 identifies the essential buses for MNGP. Essential buses 15 and 16 are the MNGP emergency buses. 55 SYSTEM MALFUNCTIONS SSS: INITIATING CONDITIONS NEI 99-01 Re v 6 Monticello N ucl e ar Generatin g Plant Inability to sh u t down the reactor causing a cha ll enge to (core cooling [PWR] Inabi l ity to s h u tdown the reactor ca u s ing a challenge to RPV water l eve l or I RP V water leve l [BWR]) o r R CS h eat r emoval. RCS h ea t remova l. Difference I Deviation I Justification Non e THRESHOLDS NEI 99-01 Rev 6 Monticello Nucl e ar G e nerating Plant (1) a. An automatic or manual (trip [PW R] I sc ram [BWR]) did SSS.I a. A n a u tomatic or man u al scram did not reduce re ac tor ower not s h ut down t he rea c tor. to le ss tha n 4%. A ND AN D b. A ll m an u a l act i o n s to s hu t d own t h e reactor are not<S uccessful in rcduoing reactor ower to les s than 4%. b. A ll man u a l actions to sh u t down the reacto r have AN D bee n un s u ccessfu l. c. EITHER of the following co n d i tion s ex i st: A ND

Reactor vesse l water level cannot be re.stored and c. EITH E R of the following conditions exist: maintain e d above -1 49" (Minimum Steam Cooling RPV * (Site-s pecific in dication of an i n a bil ity to adeq u ate l y Wat e r Level) remove hea t from t h e co r e)

hleat Ca a o ity bi miJ (IHCL) exceeded * (S i te-s pec i fic i ndicat i on of an i na bil i t y to adequa t e l y remove h ea t from t h e R CS) Differ e n c e I D ev i a tion I Justification 56 SYSTEM MALFUNCTIONS Difference:

Site specific information provided for SSS.la and SSS.lb. Reactor power level indicative of a successful shutdown is identified in C.S.1-2007, Failure To SCRAM. Justification:

MNGP EOP guidance -reactor power less than 4% is indicative of a successful shutdown of the reactor. NEI 99-01Rev6 Developer Note guidance states that the use of EOP indication of successful shutdown is acceptable.

Difference:

Site specific information provided for SSS.l.c. Justification:

Reactor vessel level of -149" is indicative of a water level that is unable to adequately remove heat from the core (EOP C.S.1-1100, Part J). Exceeding the HCL is an indication of the inability to adequately remove heat from the RCS (EOP C.5.1-1200, Part E). 57 SYSTEM MALFUNCTIONS SSS: INITIATING CONDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Lo ss of a ll Vital DC power for 15 minute s or l o n g er. Lo ss of all Yit a i DC power for 15 minute s or lon g er. Difference I Deviation I Justification None THRESHOLDS NEI 99-01 R ev 6 Monticello Nuclear Generating Plant (J) Indi ca t e d vo lt age i s l ess th a n (s it e-s p ec ifi c bu s vo l tage va lu e) o n SS 8.l Ind i c a t e d vo lt age i s l ess th a n J 10 DC o n ALL 125 VDC Vit a l DC ALL (s it e-s p e cific Vit a l DC bu s e s) for 15 minut es or l o n ge r. bu s e s for 15 minut es or l o n ge r. Difference I De v iation I Justification Difference:

S ite s pecific information pro vi d e d. Justification:

The 110 VDC threshold is based on an average for both Division I and II batteries for battery voltages at 15 minutes prior to reaching the minimum required terminal voltage for equipment operation. Calculations CA-02-179 and CA-02-192 provide the basis for the selected voltage. USAR-08.05 supports the use of 125 VDC Vital DC buses for this threshold.

58 j SYSTEM MALFUNCTIONS SAl: IN I TIATING CON D ITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Loss of all but one AC power so urce to emergency buse s for l 5 mjnutes or Lo ss of a ll but one AC power so urce to e s sentia l bu ses for 15 mjnute s or l o n ger. l onger. Difference I Deviation I Justification Difference:

NEI 99-01 Rev 6 IC refers to emergency bu ses. MNGP IC r efers to esse nti a l buse s. J ustifica t i on: Termjnology difference

-the essential buses at MNGP are the emergency buses. THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) a. AC power capabil it y to (s it e-spec ifi c emergency buses) is SAl.l a. AC power capab ilit y to essent i a l bu s es 15 and 16 i s reduced to reduced to a s ingle power so urce for 15 mjnutes or longer. a si ngle power so urce (Table S 1) fo r 15 mjnutes or lon ger. AND AND b. Any addi ti o n al sing l e power so ur ce fa ilu re will r esult in a l oss b. Any additio n a l s in g l e power so ur ce fa ilur e will result in a of al l AC power to SAFE TY SYSTEMS. l oss of a ll AC power t o SAFETY SYSTEMS. Table Sil l lR R eserve lAR Reserve Transformeu 2J R Auxiliary Transformer

11 Emergency Diesel Generator

1 12 Emergency Die s e l Generator Difference I Deviation I Justification Difference:

Site s pecific information provided.

Justification:

Table Sl incorporates EPFAQ 2015-015 guidance.

Ops Man B.09.06-01 identifies the essential buses for MNGP. Terminology difference

-the essential buses at MNGP are the emergency buses. 59 SYSTEM MALFUNCTIONS NEI 99-01 Rev 6 UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.

Difference I Deviation I Justification None NEl 99-01 Rev 6 Monticello Nuclear Generating Plant UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.

60 SYSTEM MALFUNCTIONS (1) a. An UNPLANNED event results in the inability to monitor SA2.1 a. An UNPLANNED event results in the inability to monitor one one or more of the following parameters from within the or more of the following parameters from within the Control Control Room for 15 minutes or longer. Room for 15 minutes or longer. Reactor Power [BWR parameter list] [PWR parameter list] RPV Water Level Reactor Power Reactor Power RPV Pressure Primary Containment Pressure RPV Water Level RCS Level Suppression Pool Level RPV Pressure RCS Pressure Suppression Pool Temperature Primary Containment In-Core/Core Exit AND Pressure Temperature

b. ANY of the following transient events in progress.

Suppression Pool Level Levels in at least (site-* Automatic or manual runback greater than 25% thermal specific number) steam reactor power generators

Electrical load rejection greater than 25% full electiical Suppression Pool Steam Generator Auxiliary load Temperature or Emergency Feed Water

Reactor scram Flow

ECCS actuation

Thermal power oscillations greater than 10% AND b. ANY of the following transient events in progress.

Automatic or manual runback greater than 25% thermal reactor power

Electrical load rejection greater than 25% full electrical load

Reactor scram [BWR] I trip [PWR]

ECCS (SI) actuation

Thermal power oscillations greater than 10% [BWR] Difference I Deviation I Justification

' None 61 SYSTEM MALFUNCTIONS SAS: INIT I ATING C O NDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant A ut o m a tic o r manu a l (tr ip [PWR] I sc r a m [BWR]) fa il s to s hut d ow n th e A ut o m a t ic or m a nu a l s c ra m fa il s to s hutd ow n th e r eac t or, a nd s ub se qu e nt r eac t o r, a nd s ub se qu e nt m a nu a l ac ti o n s take n a t th e r eac t o r co ntr o l co n so l es m a nual ac ti o n s t ake n a t th e main co ntrol board s a r e n o t s uc cess ful in s huttin g are n o t s u ccess ful in s huttin g d ow n th e r eac t o r. d o wn th e r eac t o r. Difference I Deviation I Justification Difference:

Monticello IC replaces 'reactor contro l consoles' with 'main control boards'. Justification:

Human factors consideration

-Monticello personnel recognize

'main control boards' as being 'reactor control consoles'.

Change does not affect the intent of the NEI 99-01 Rev 6 IC. THRESHOLDS NEI 99-01Rev6 Monticello Nuclear Generating Plant (1) a. An a ut o m a t ic or m a nu a l (trip [PWR] I scra m [BWR]) did SAS.I a. An a utom a tic or m a nu a l sc ram did n ot r e duce react o r owe r to n o t s hutd ow n the r eac t o r. l ess than 4%. AND AND b. M a nu a l ac ti o n s t ake n a t th e main contro l boards a r e n o t s u ccess ful in reducing r eac tor ower to l ess v h an 4%. b. M a nu a l ac ti o n s t ake n a t th e r eac t o r co n tro l co n so l es are n o t s u ccess ful i n s huttin g d o wn th e r eac t or. Difference I Deviation I Justification Difference:

NEI 99-01Rev6 Thre s hold (l)b. refer s to 'r e actor control con s ole s'. MNGP Threshold SAS.lb. s pecifies 'main control board s'. Justification:

Human Factors consideration

-the main control boards are easily recognized by Monticello personnel as the location of the 'reactor control consoles'.

Difference:

Site specific information provided for SAS.la. Reactor power leve l indicative of a successful shutdown is identified in C.S.1-2007, Failure To SCRAM. Justification:

MNGP EOP guidance -reactor power less than 4% is indicative of a successful shutdown of the reactor. NEI 99-01 Rev 6 Developer Note guidance states that the use of EOP indication of successful shutdown is acceptable.

62 SYSTEM MALFUNCTIONS SA9: IN IT IATING CON D ITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant Hazardous event affecting a SAFETY SYSTEM needed for the current Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. operating mode. Difference I Deviation I Justification None THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (l) a. The occurrence of ANY of the following hazardous events: SA9.l a. The occurrence of ANY of the following hazardous events:

Seismic event (earthquake)

Seismic event (Earthquake)

Internal or externa l flooding event

Internal or external flooding event

High winds or tornado s trike

High winds or torn a do strike

FIRE

FIRE

EXPLOSION

EXPLOSION

(site-s pecific hazards)

Ri ve r l eve l g r ea t e r th a n 9 1 9 ft e l

Other events with simi l a r hazard characteristics as

Ri ver l evel l ess th a n 90 0.5 f t c l determined by the Shjft Manager

Other events with s imilar haz a rd characteristics as AND determined by the Shift Manager b. EITHER of the following:

AND l. Event damage ha s caused indication s of degraded b. EITHER of the following:

performance in at lea s t one train of a SAFETY SYSTEM need e d for the current operating mod e.

Event dama ge h as ca u se d indi ca tion s of degraded OR performance in a t le as t one train of a SAFETY SYSTEM 2. The event has caused VISIBLE DAMAGE to a needed for the cwTent operating mode. SAFETY SYSTEM component or s tructure

The event has caused VISIBLE DAMAGE to a SAFETY needed for the current operat in g mode. SYSTEM compone nt or structure needed for the current operating mode. Difference I Deviation I Justification 63 SYSTEM MALFUNCTIONS Difference:

MNGP Threshold SA9.lb. -deleted OR and replaced numbered indicators with bulleted listing. Justification:

Editorial revision -consistent use of the nesting within the EAL thresholds.

Intent of NEI 99-01 Rev 6 threshold remains satisfied.

Difference:

Site specific information provided.

Justification:

Seismic indication is identified in ARP C.6-006-C-08 and C.6-006-C-13.

MNGP procedure A.6 identifies the criteria for high and low river levels. 64 SYSTEM MALFUNCTIONS SUl: INITIATING CONDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating P l ant Loss of all off s ite AC power capability to emerge n cy bu s e s for 15 minutes or Lo s s of all off s ite AC power capab ilit y to esse nti a l bu ses for 15 minute s or longer. l onger. Difference I Deviation I Justification Difference:

NEI 99-01 Rev 6 IC refers to emergency bus es. MNGP IC refers to esse nti a l bu s e s. Justification:

Terminology difference

-the essential buses at MNGP are the emergency buses. THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating Plant (1) Loss of ALL off s ite AC power capabi li ty to (s ite-s pecific SUl.l Loss of ALL offs it e AC power capability (T a ble S2) to esse nti a l emergency bu s e s) for 15 minute s or l o n ger. bu ses LS a nd 16 for 1 5 minute s or l onger. Tab l e S2 IR Re se r ve T ra n s form e r I AR Re se rve T ra n s form e r 2R Auxili a r y Tran s former Difference I Deviation I Justification Difference:

Site s pecific information provided.

Justification:

Table S2 incorporates EPF AQ 2015-015 guidance.

Ops Man B.09.06-01 identifies the essential buses for MNGP. Terminology difference

-the essential buses at MNGP are the emergency buses. 65 SYSTEM MALFUNCTIONS

.*

INITIATING NEI 99-01 Rev 6 Monticello Nuclear Generating Plant UNPLANNED loss of Control Room indications for 15 minutes or longer. UNPLANNED loss of Control Room indications for 15 minutes or longer. Difference I Deviation I JustificatiOn

. None THRESHOLDS NEI 99-01 Rev 6 (1) a. **. An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer. [BWR parameter list] Reactor Power RPV Water Level RPV Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature

[PWR parameter list] Reactor Power RCS Level RCS Pressure In-Core/Core Exit Temperature Levels in at least specific number) steam generators Steam Generator Auxiliary or Emergency Feed Water Flow Difference l Deviation I Justification None Monticello Nuclear Generating Plant SU2.1 a. An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer. Reactor Power RPV Water Level RPV Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temoerature . 66 SYSTEM MALFUNCTIONS SU3: INITIATING C O NDITI O NS NEI 99-01 Rev 6 Monticello N u clear Generating P l ant Reactor coo lant activity greater than Technical Specification allowable limits. R eactor coo l ant activity greater than Technical Specification allowable Limits. Difference I Deviation I Justification None THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating P l ant ( 1) (Site-specific radiat i o n monitor) reading greater than (site-specific SU3.l Offgas Pretreatment Radiation Monitor (RM-1 7-1 SOA or RM-17-value). !SOB) high radiation a l arm (4-A-12) received. (2) Sample a n a l ysis indicates that a reactor coo l ant activity value is SU3.2 Coolant sam le activity grea ter than 0.2 µCi/gm dose equivalent I-greate r than an a ll owab l e limit spec ifi ed in Technical Specifications.

1 31. Difference I Deviation I Justification Difference:

Site specific information provided.

Justification:

Information supporting EAL thresholds SU3.1 and SU3.2 is included in TS 3.4.6, TS 3.7.6, ARP C.6-004-A-12, and Ops Man B.05-11-01.

67 SYSTEM MALFUNCTIONS SU4: INITIATING CONDITIONS NEI 99-01 Rev 6 Monticello Nuclear Generating P l ant RCS l eakage for 15 minute s or l onger. RCS l eakage for 1 5 minute s or longer. Difference I Deviation I Justification None THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generat i ng P l ant (1) RCS unid e nti fied or pressure boundary l eakage grea ter than (site-SU4.l RCS unid entified or pre ss ure boundary l eakage greate r than I 0 g m specific value) for 15 minute s or longer. for 15 minutes or l onger. (2) RCS identified leaka ge greater than (s ite-specific value) for 15 minutes or l o n ge r. SU4.2 RCS identified leak age greater than 2 5 g m for 15 minute s or (3) Leakage from the RCS to a l ocatio n o ut side co n tai nm ent greater l onger. than 25 gp m for 1 5 minutes or longer. SU4.3 Leakage from th e RCS to a location outside rim a r y contai nm e nt grea t er than 25 gpm for 15 minutes or longer. Difference I Deviation I Justification Difference:

Site specific information provided for RCS leakage per NEI 99-01 Rev 6 guidance.

Difference:

MNGP SU4.3 EAL threshold includes 'primary' as a further clarifier for leakage outside containment.

Justification:

NSSS (BWR) terminology clarification that does not affect the classification of the event. 68 SYSTEM MALFUNCTIONS SUS: INITIATING C O NDITIONS NEI 99-01Rev6 Montice ll o Nuclear Generating P l ant A u to m a ti c or m an u al (tr i p [PWR] I scram [BWR]) fails t o s hutd ow n the A u tomat i c or man u al scra m fai l s to sh u t d ow n th e reactor. r eac t or. Difference I Deviation I Justification None THRESHOLDS NEI 99-01 Rev 6 Mont i cello Nuclear Generating P l ant (I) a. An a ut o m at ic (trip [PWR] I scra m [BWR]) d i d not SUS.I a. An initial a ut o m at i c or manual sc r am d id n o t reduce reactor s hutd ow n t h e reactor. ower to less than 4%. AND AND b. A su b se qu e nt ma nu a l act i o n take n a t t h e r eactor co n tro l b. ANY of the following is s ucce ss ful in reducing reactor ower co n so l es is s u ccessfu l in s hu tt in g d ow n t h e r eac t or. to less than 4%: (2) a. A m a nu a l trip ([PWR] I scra m [BWR]) did n ot s hutd ow n

Manual scram pushbutton s th e r ea ct or.

Mode switch to Shutdown AND EITHER of th e fo ll ow in g:

Alternate Rod Insertion (AR I) b.

Subse uent automatic scram 1. A s ub se qu e nt m a nu a l act i on t ake n a t t h e reac t o r co ntr o l co n so l es i s s u ccess ful in s hu tt in g dow n th e reactor. OR 2. A s u bseq u en t a u to m a ti c (t r ip [PWR] I scra m [BWR]) is s u ccessfu l in s huttin g dow n th e reactor. Difference I Deviation I Justification 69 SYSTEM MALFUNCTIONS Difference:

NEI 99-01, Rev 6 guidance does not include the clarifier

'initial' in EAL thresholds (1) or (2). Justification:

MNGP EAL threshold SUS.1.a adds the clarifier

'initial' to provide additional guidance that an initial trip signal (either automatic or manual) was not successful.

Credit is then allowed for either a subsequent automatic trip signal or manual action to successfully trip the reactor. In this situation, the RPS or operator action succeeds in shutting down the reactor. Difference:

Justification:

Difference:

Site specific information provided.

Reactor power level indicative of a successful shutdown is identified in MNGP EOP guidance.

Reactor power of less than 4% is indicative of a successful shutdown of the reactor. NEI 99-01Rev6 Developer Note guidance states that the use of EOP indication of successful shutdown is acceptable.

Therefore, the intent of NEI 99-01 Rev 6 threshold is still satisfied.

NEI 99-01 Rev 6 utilizes two EAL thresholds, one for automatic trip and one for manual trip. MNGP combined these thresholds into one threshold (SUS.1). Justification:

Human factors consideration.

The EAL threshold was revised to facilitate determination by the decision maker that the conditions have been satisfied.

Even though these thresholds are combined, the intent of NEI 99-01 Rev 6 EAL threshold remains satisfied.

Difference:

NEI 99-01 Rev 6 EAL Threshold (2)b uses numbers to identify the two, threshold conditions and does not included manual actions that are initiated by operators to trip the reactor. Justification:

Editorial revision for clarity. This revision does not affect the EAL threshold and is consistent with the format of other EAL thresholds.

In addition, MNGP EAL threshold SUS.l.b identifies the manual actions taken by the reactor operator at the main control boards to initiate a reactor trip. 70 SYSTEM MALFUNCTIONS SU6: INITIATING C O NDITIONS NEI 99-01 Rev 6 Montic e llo Nuclear Generatin g Plant L oss o f a ll o n site or offsite comm u nications capa b i lit ies. L oss of all o n si te o r offsi t e com mu nications ca p a b i li ties. Difference I De v iation/ Justification None THRESHOLDS NEI 99-01 Rev 6 Monticello Nuclear Generating P l ant (1) Loss of ALL of t h e fo ll ow in g o n s i te comm un ica t io n m et h ods: SU6.l Loss of A LL of th e fo ll ow in g ons it e co mmu nicatio n met h o d s: (s it e-specific li s t of comm un ica t io n s met h o d s)

Commercial Tele hones (2) L oss of ALL of th e fo ll ow in g ORO comm uni ca t ions me t hods:

Plant Telephone s (s it e-spec i fic li s t of co mmun ications m e th o d s)

Portable radio s (3) L oss of ALL of t h e fo ll ow in g NRC co m m un icatio n s methods: (si t e-specific l ist of comm un ica t io n s me th ods)

Plant PA System SU6.2 Loss of A LL of th e fo ll ow in g Offsite R espo n se Or ga ni zation (ORO) co mmuni ca ti o n s m e th o d s:

Commercia l Te l ephones

Direct Dedicated Telephones

Radio/Receiver Transmitter SU6.3 Loss of ALL of th e fo ll ow in g N R C comm u n i cat i o n s met h ods:

Federal Telecommunications System (FTS)

Commercial Tele hones Difference I De v iation I Justification Diff e rence: S i te s pecific information provided.

71 L-MT-17-012 NSPM ATTACHMENT 2 MONTICELLO NUCLEAR GENERATING PLANT License Amendment Request to Revise the Emergency Action Level Scheme Emergency Action Level Technical Bases Document (Red-Line Version) (For Information Only) (154 pages to follow)

MONTICELLO NUCLEAR GENERATING PLANT EMERGENCY ACTION LEVELS INITIATING CONDITIONS, THRESHOLD VALUES, AND BASES TABLE OF CONTENTS 1 REGULATORY BACKGROUND

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

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

.... 1 1.1 O PERATING REACTO R S **********************..*..*.*...*..................

- - ..............*

- - - - ..**.*..*.......

..*..*. 1 1.2 I NDEPE ND EN T SPEN T FUE L S TORAG E I NSTALLATI O N (ISFSl) .**.....**......*.................*..

1 1.3 N RC ORD E R E A-12-051 ....................*

.......*..**..........

.******...*.*.*.*****.*..*..*..***.*................ 2 2 KEY TERMINOLOGY USED IN NEI 99-01 .......................

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

4 2.1 E ME R GENCY C LASS I F I CATI O N LEVEL (E CL) ............................................................

.. .4 2.2 I NIT I AT I NG CO NDIT I ON (IC) .................

...****.*.....*.....*************.*****.***************.***************** 6 2.3 EMERGEN C Y ACTI O N L EVEL (EAL) .*.......*.....................*................*..

..*...*.**..*....*..*

- - - - 6 2.4 FI SSION PROD UCT B A RRI E R THRESH O L D *.*...........**.........

....*.......

.***************.****.********* 6 3 DESIGN OF THE NEI 99*01 EMERGENCY CLASSIFICATION SCHEME ADOPTED BY MNGP ...............................................................................

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

8 3.1 ASSIGNMENT O F EME R GENCY CLASS I FICATI O N LEVELS (EC L S) ..*.......***....*............. 8 3.2 T YPES O F I N I TIATING CO ND I T IO NS AN D EME R GENCY ACTI O N LEVELS ....................

11 3.3 NSSS-MNGP SPECIFIC D ES I GN DIFFERENCES CONSIDERATIONS

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

11 3.4 OR GAN I ZA TIO N AN D PR ESENTATION O F GENE RI C I NFORMATI O N ....*......

- ..........*..*.

1 2 3.5 IC AN D EA L M O DE APPL I CA BI L I TY ***.******.****...............

- ..**......

......*******************..*..... 13 4 SITE-SPECIFIC SCHEME DEVELOPMENT GUIDANCE ..............................................

16 4.1 G ENE RA L I MP L EMENTAT IO N GU ID ANCE *..*..***.**........**....*.......*.......*......***...........*..** 1 6 4.2 CRI TICAL C HARACTE RI STICS .....*......................................................*......

................... 1 7 4.3 I NST R UMENTATI O N USE D F OR EA L S ..................

..*********.****.**..........................

.......*... 18 4.4 P RESENTAT IO N O F SC H EME I NF OR MAT IO N T O USE R S ..***...*......*.*..............***............ 18 4.5 I NTEGRA TIO N O F ICs/E A Ls W I T H PL ANT PRO CE D URES ....*......*........**....

........***.**.* 20 4.6 B ASIS DO CUMENT ..................................................

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

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

20 4. 7 EA L/THR ES HO LD REFE R ENCES T O A OP AN D E OP SETP OI NTS/CR ITE R IA ..............

21 4.8 D EVEL OP E R AND USER F EE DB AC K ........................

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

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

......... 2 1 5 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS

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

22 5.1 G ENERAL CO NSIDERA TIO NS *******.............................

..*.**.************..*........*..........*..........

22 5.2 C LASS I F I CA TIO N ME THODOLO GY ...................................................

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

23 5.3 CL ASS I F I CA TIO N OF M U L TIPLE EVENTS AN D CO NDITIONS ...........

............................. 23 5.4 C O NS ID ERATI O N O F M OD E C HANGES D URING CLASS I FICATI O N ..............................

23 5.5 CL ASSIF I CAT IO N O F I MMINENT C O NDIT IO NS ........................................

..................... 2 4 5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING

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

24 5.7 CLASSIFICATION OF SHO RT-LI VED EVENTS ..............................

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

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

25 5.8 CLASSIFICATION OF TRANSIENT CONDITIONS

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

25 5.9 AFTER-THE-F ACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION

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

26 5.10 RETRACTION OF AN EMERGENCY DECLARATION

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

26 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS *******.****.***********

27 7 COLD SHUTDOWN/

REFUELING SYSTEM MALFUNCTION ICS/EALS ************..*.*.*.

45 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS *********

70 88 9 FISSION PRODUCT BARRIER ICS/EALS *************.***************..*****..***.**************.****

73+.l 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS **** 89 S1 11 SYSTEM MALFUNCTION ICS/EALS ***************************************.*************

- - - - 115"2 APPENDIX A -ACRONYMS AND ABBREVIATIONS

- - - - ....**.

A*1 APPENDIX B -DEFINITIONS

8-1 II DEVELOPMENT OF EMERGENCY AC TI ON L EVELS FOR NON REACTORS MONTICELLO NUCLEAR GENERATING PLANT 1 REGULATORY BACKGROUN D 1.1 O PERATING REACTORS T i t l e 10, Code of Federa l Regu l ations (CPR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulat i o n s that apply to nuclear power facil i ties. Several of t h ese regulations govern various aspects of an emergency classification scheme. A rev i ew of the re l evant sect i ons listed be l ow will aid the reader in u nderstanding the key termino l ogy provided in Section 3.0 of th i s document.

10 CPR§ 50.47(a)(l)(i)

10 CPR§ 50.47(b)(4)

10 CPR§ 50.54(q)

10 CPR§ 50.72(a)

10 CPR § 50, Appendix E, IV .B, Assessment Actions

10 CPR§ 50, Appendix E, IV.C, Activation of Emergency Organization The above regu l ations are s u pp l eme n ted by various reg u latory guidance docume n ts. Three documents of partic ul ar re l evance to NEI 99-01 are:

NUREG-0654/FEMA-REP-l, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants]

NUREG-1022, Event Reporting Guidelines 10 CFR § 50. 72 and§ 50. 73

Regu l atory Guide 1.10 1 , Emergency Response Planning and Preparedness for Nuclear Power Reactors The above list is not all inolusive and it is strongly reoommended that soheme developers eonsult with licensing/regulatory compliance personnel to identify and understand all applicable requirements and guidance.

Questions may also be directed to the NEI Emergency Preparedness staff. 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) Se l ected guidance in NEI 99-01 is app l icab l e to licensees electing to use their 10 Gf'R -:W the Monticello Nuclear Generating Plant (MNGP) emergency p l an to fu l fi ll t h e re qui rements of 10 CPR 72.32 for a stand-a l one ISFSI. The emergency classificat i on l eve l s applicab l e to an ISFSI are consiste n t with the requirements of 10 CPR § 50 and the guidance in NUREG 0654/FEMA-REP-l. The initiati n g condit i ons germane to a 10 CPR§ 72.32 emergency p l an (as descr i bed in NUREG-1567) are subsumed wit hi n the c l assification scheme for a 10 CPR§ 50.47 emergency p l an.

The generic MNGP IC s and EA Ls for an ISFSI are presented in Section 8 , ISFSI ICs/EALs.

IC E-H Ul covers the spectrum of credible natural and man-made events included within the scope of an-the ISFSI design. This IC is not applicable to installations or facilities that may process and/or repackage spent fuel (e.g., a Monitored RetrieYable Storage Pacility or an fSf'ST at a spent fuel processing facility).

In addition , appropriate aspects of IC HUI and IC HAI should also be included to address a HOSTILE ACTION directed against an-the ISFSI. The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-1140 , A Regulator y Analysis on Emergenc y Preparedness for Fuel C ycle and Other Radioactive Mat erial Licensee

s. NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public hea l th and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed l rem Effective Dose Equivalent.

Regarding the above information , the expectations for an offsite response to an Alert classified under a 10 CFR § 72.32 emergency plan are generally consistent with those for a Notification of Unusual Event in a 10 CFR § 50.47 emergency plan (e.g., to provide assistance if requested).

Also, the licensee's MNGP Emergency Response Organization (ERO) required for a 10 CFR § 72.32 emergency plan is different than that prescribed for a 10 CFR § 50.47 emergency plan (e.g., no emergency technical support function).

1.3 NRC ORDER EA-12-051 The Fukushima Daiichi accident of March 11 , was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity , and ultimately led to core damage in three reactors.

While the loss of power also impaired the spent fuel pool cooling function , sufficient water inventory was maintained in the pools to preclude fuel damage from the loss of cooling. Following a review of the Fukushima Daiichi accident , the NRC concluded that several measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule , 10 CFR 50.109(a)(4)(ii).

Among them was to provide each spent fuel pool with reliable level instrumentation to significantly enhance the ability of key decision-makers to allocate resources effectively following a beyond design basis event. To this end , the NRC issued Order EA-12-051, Is s uance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumen t ation , on March 12 , 2012 , to all US nuclear plants with an operating license , construction permit , or combined construction and operating license. NRC Order EA-12-051 states , in part , "All licensees

... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel:

(1) level that is adequate to support operation of the normal fuel pool cooling system , (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck , and (3) level where fuel remains covered and actions to implement 2

make-up water addition should no longer be deferred." To this e nd , all licensees must provid e:

A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool;

A display in an area accessible following a severe event; and

Independent electrical power to each instrument channel and provide a n alternate remote power connection capabilit y. NEI 12-02 , Indu stry Guida nc e for Comp lianc e wit h N RC Order EA-12-051, " To Modify License s with Regard to Reliabl e Spe nt Fuel Pool In strumentation'

', provides guidance for complying with NRC Order EA-12-051.

}JEJ 99 01 , Revision 6 , This document includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-051. These EALs are included within existing IC and new ICs A82-RS2 and AG2-RG2. Associated EAL notes , eases and de't'eloper notes are also pro*,,ided.

It is recommended that these EAbs 0e implemented when tl'le enhanced spent fuel pool level instn:1mentation is availaele for l:ISe. The regl:llatory process that licensees follow to make changes to their emergency plan , including non scheme changes to EAbs , is I 0 Cf<R 50.54 (q). In accordance

.vith this regulation , licensees are responsiele for e*1aluating a proposed change and determining 1.vhether or not it results in a reduction in the effectiveness of the plan. As a resl:llt of the licensee's determination , the licensee will either make the cha1'lge or Sl:lemit it to the NRG for prior review and approval in accordance with 10 Cf<R 50.90. 3 L __ 2 MNGP KEY TERMINOLOGY USED IN NEI 88 01 There are several key terms that appear throughout the NET 99 Ol emergency classification methodology for MNGP. These terms are introduced in this section to support understanding of subsequent material.

As an aid to the reader , the following table is provided as an overview to illustrate the relationship of the terms to each other. Emergency Classification Level Um:1sual E>.*ent GE I A!et:t SAE I SAE Alert I @Unusual Event Initiating Condition Initiating Condition Initiating Condition Initiating Condition Emergency Action Emergency Action Emergency Action Emergency Action Level (1) Level (1) Level (1) Level (1)

Operating Mode

Operating Mode

Operating Mode

Operating Mode Applicability Applicability Applicability Applicability

Notes

Notes

Notes

Notes

Basis

Basis

Basis

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

This includes the Emergency Action Level (EAL) p l us the associated Operating Mode Applicability, Notes and the informing Basis information.

In the Recognition Category F matrices , EALs are referred to as Fission Product Barrier Thresholds

the thresholds serve the same function as an EAL. 2.1 EMERGENCY CLASSIFICATION LEVEL (ECL) One of a set of name s 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) resu l ting onsite and offsite response actions. The emergency classification leve l s, in ascending order of severity , are

Notification of Unusua l Event (N G UE)

Alert

Site Area Emergency (SAE)

General Emergency (GE) 2.1.1 Notification of Unusual Event (N G UE)+ Events are in progress or have occurred which indicate a potential degradation of the l evel of safety of the plant or indicate a security threat to facility protection has been +This term is sometin:ies shortened to Unusua l El*ient (UEI) or other simi l ar site speeifie tern1inology.

The tenm Notifieation of Unusual e't'ent , }-JQUE and Unusua l Elvent are used interehangeably througl:im1t this 4

initiated.

o releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs. Purpose: The purpose of this classification is to assure that the first step in future response has been carried out, to bring the operations staff to a state of readiness , and to provide systematic handling of unusual event information and decision-making.

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

2.1.4 General Emergency (GE) Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.

Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Purpose: The purpose of the General Emergency declaration is to initiate predetermined protective actions for the public , to provide continuous assessment of information from the licensee and offsite organizational measurements , to initiate additional measures as indicated by actual or potential releases , to provide consultation with offsite authorities , and to provide updates for the public through government authorities. 5 2.2 INITIATING CONDITION (IC) An event or condition that aligns with the definition of one of the four emergency classification l evels by virt ue of the potential or actua l effects or consequences.

Discussion:

An IC describes an event or condition , the severity or consequences of which meets the definition of an emergency classification level. An IC can be expressed as a continuous , measurable parameter (e.g., RCS leakage), an event (e.g., an earthquake) or the status of one or more fission product barriers (e.g., loss of the RCS barrier).

Appendix 1 of NUREG-0654 does not contain example Emergency Action Levels (EALs) for each ECL , but rather Initiat in g Conditions (i.e., plant conditions that indicate that a radiological emergency , or events that could l ead to a radio l ogical emergency , has occurred).

NUREG-0654 states that the Initiating Conditions form the basis for establishment by a licensee of the specific plant instrumentation readings (as applicable) which , if exceeded , would initiate the emergency classification.

Thus , it is the specific instrument readings that would be the EALs. Considerations for the assignment of a particular Initiating Condit i on to an emergency classification l evel are discussed in Section 3. 2.3 EMERGENCY ACTION LEVEL (EAL) A pre-determined, site-specific, observab l e threshold for an Initiating Condition that , when met or exceeded , places the plant in a given emergency classification level. Discus sion: EAL statements may utilize a variety of criteria including instrument readings and status indications; observable events; results of calculations and analyses; entry into particular procedures; and the occurrence of natural phenomena.

2.4 FISSION PRODUCT BARRIER THRESHOLD A pre-determined , site-specific , observab l e threshold indicating the l oss or potential loss of a fission product barrier. Discussion:

Fission product barrier thresho ld s represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment.

This concept relies on multiple physical barriers , any one of which , if maintained intact , precludes the release of significant amounts of radioactive fission products to the environment.

The primary fission product barriers are:

Fuel Clad

Reactor Coo lan t System (RCS)

Containment Upon determination that one or more fission product barrier thresholds have been exceeded , the combination of barrier loss and/or potential loss thresholds is compared to the fission product barrier IC/EAL criteria to determine the appropriate ECL. In some accident sequences, the ICs and EALs presented in the Abnormal Radiation 6

Levels/ Radiological Effluent (A R) Recognition Category will be exceeded at the same time, or shortly after, the loss of one or more fission product barriers.

This redundancy is intentional as the former ICs address radioactivity releases that result in certain offsite doses from whatever cause , including events that might not be fully encompassed by fission product barriers (e.g., spent fuel pool accidents , design containment leakage following a LOCA , etc.). 7 3 DESIGN OF THE NEI 99-01 EMERGENCY CLASSIFICATION SCHEME ADOPTED BY MNGP 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS) An effective emergency classification scheme must incorporate a realistic and accurate assessment of risk , both to plant workers and the public. There are obvious health and safety risks in underestimating the po te ntial or actual threat from an event or condition; however , there are also risks in overestimating the threat as well (e.g., harm that may occur during an evacuation).

The NEI 99-01 emergency classification scheme attempts to strike an appropriate balance between reasonably anticipated event or condition consequences , potential accident trajectories , and risk avoidance or minimization.

NSPM has adopted the NEI 99-01 scheme , adding site-specific information as appropriate. This section discusses the background for deve l opment of the N E I 99-01 scheme and adds MNGP specific details where appropriate. There are a range of " non-emergency events" reported to the US Nuclear Regulatory Commi s sion (NRC) staff in accordance with the requirements of 10 CPR§ 50.72. Guidance concerning these reporting requirements , and example events , are provided in NUREG-1022.

Certain events reportable under the provisions of 10 CFR § 50.72 may also require the declaration of an emergency.

In order to align each Initiating Conditions (IC) with the appropriate ECL , it was necessary to determine the attributes of each ECL. The goal of this process is to answer the question , " What events or conditions should be placed under each ECL ?" The following sources provided information and context for the development of ECL attributes.

Assessments of the effects and consequences of different types of events and conditions

Typical abnormal and emergency operating procedure setpoints and transition criteria

Typical Technical Specification limits and controls

Radiological Effluent Technical Specifications (RETS)/Off s ite Dose Calculation Manual (ODCM) radiological release limits

Review of selected Updated Final Safety Analysis Report (UFSAR) accident analyses

Environmental Protection Agency (EPA) Protective Action Guidelines (PAGs)

NUREG 0654 , Appendix 1 , Emergency Action Level Guideline s for N uclear Power Plant s

Industry Operating Experience

Input from industry subject matter experts and NRC staff members The following ECL attributes were created by the Revision 6 Preparation Team to aid in the development of I Cs and Emergency Action Levels (EALs). The se team deeided to inelude the attributes in this revision sinee they may be useful in briefing and training settings (e.g., helping an Emergency Director understand why a particular condition is classified as an Alert). Tt should be stressed that developers not attempt to redefine these 8 attrib1:1tes or apply them in any fashion that would change the generic g1:1idance contained in The attributes of each ECL are presented below. 3.1.1 Notification of Unusual Event (N Q UE) A Notification of Unusual Event , as defined in section 2.1.1 , includes but is not limited to an event or condition that involves: (A) A precursor to a more significant event or condition. (B) A minor loss of control of radioactive materials or the ability to control radiation levels within the plant. (C) A consequence otherwise significant enough to warrant notification to local , State and Federal authorities.

3.1.2 Alert An Alert , as defined in section 2.1.2 , includes but is not limited to an event or condition that involves: (A)A loss or potential loss of either the fuel clad or Reactor Coolant System (RCS) fission product barrier. (B) An event or condition that significantly reduces the margin to a loss or potential loss of the fuel clad or RCS fission product barrier. (C) A significant loss of control of radioactive materials resulting in an inability to control radiation levels within the plant , or a release of radioactive materials to the environment that could result in doses greater than 1 % of an EPA PAG at or beyond the site boundary. (D)A HOSTILE ACTION occurring within the OWNER CONTROLLED AREA , including those directed at an Independent Spent Fuel Storage Installation (ISFSI). 3.1.3 Site Area Emergency (SAE) A Site Area Emergency , as defined in section 2.1.3 , includes but is not limited to an event or condition that involves: (A) A loss or potential loss of any two fission product barriers -fuel clad , RCS and/or containment. use efeCb attributes is at the diseretieR efa lieeRsee aREI is Rat a require1'l'!eRt eftl:ie l>IRC. lfa lieeRsee el:ieoses iR iReerperate ti-le eCb attrib1:1tes iRte tl:ieir basis deeumeRt , it must be very elear ti-lat the NRG staff has Rat eRdersed their aeeeptability er applieatioR fur aRy purpose. In partie1:1lar , the staff does Rot eonsider attribute state1'l'leRts te s1:1persede ti-le establisl:ied eCb defiRitieRs.

As a res1:1lt , the use efthe attributes as a basis for justifyiRg eAb ehaRges is URaeeeptable.

9 (B) A precursor event or condition that may lead to the loss or potential loss of multiple fission product barriers within a relatively short period of time. Precursor events and conditions of this type include those that challenge the monitoring and/or control of multiple safety systems. (C) A release of radioactive materials to the environment that could result in doses greater than 10% of an EPA PAG at or beyond the site boundary. (D)A HOSTILE ACTION occurring within the plant PROTECTED AREA. 3.1.4 General Emergency (GE) A General Emergency, as defined in section 2.1.4 , includes but is not limited to an event or condition that involves: (A) Loss of any two fission product barriers AND loss or potential loss of the third barrier -fuel clad, RCS and/or containment. (B) A precursor event or condition that, unmitigated , may lead to a loss of all three fission product barriers.

Precursor events and conditions of this type include those that lead directly to core damage and loss of containment integrity. (C) A release of radioactive materials to the environment that could result in doses greater than an EPA PAG at or beyond the site boundary. (D)A HOSTILE ACTION resulting in the loss of key safety functions (reactivity control , core cooling/RPV water level or RCS heat removal) or damage to spent fuel. 3.1.5 Risk-Informed Insights Emergency preparedness is a defense-in-depth measure that is independent of the assessed risk from any particular accident sequence; however , the development of an effective emergency classification scheme can benefit from a review of risk-based assessment results. To that end, the development and assignment of certain ICs and EALs also considered insights from several site-specific probabilistic safety assessments (PSA -also known as probabilistic risk assessment , PRA). Some generic insights from this review included:

1. Accident sequences involving a prolonged loss of all AC power are significant contributors to core damage frequency at many Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). For this reason , a loss of all AC power for greater than 15 minutes, with the plant at or above Hot Shutdown, was assigned an ECL of Site Area Emergency.

Precursor events to a loss of all AC power were also included as an Unusual Event and an Alert. A station blackout coping analyses performed in response to 10 CFR § 50.63 and Regulatory Guide 1.155, Station Blackout, may be used to determine a time-based criterion to demarcate between a Site Area Emergency and a General Emergency.

The time dimension is critical to a properly anticipatory emergency declaration since the goal is to maximize the time available for State and local officials to develop and 10 implement offsite protective actions. For MNGP , the coping analysis determined that MNGP is a four (4) hour coping plant. This provides the basis for the time-based demarcation criterion between a Site Area Emergency and a General Emergency for MNGP. 2. For severe core damage events, uncertainties exist in phenomena important to accident progressions leading to containment failure. Because of these uncertainties , predicting the status of containment integrity may be difficult under seve re accident conditions.

This is why maintaining containment inte gr it y alone following sequences leading to severe core damage is an insufficient basis for not escalating to a General Emergency.

3. PSAs indicated that leading contributors to latent fatalities were sequences involving a containment bypass, a large Loss of Coolant Accident (LOCA) with early containment failure , a Station Blackout lasting longer than the site-specific coping period , and a reactor coolant pump seal failure. The generic EAL methodology needs to be sufficiently rigorous to address these sequences in a timely fashion. 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS The NEI 99-01 methodology adopted by MNGP makes use of symptom-based , based and event-based ICs and EALs. Each type is discussed below. Symptom-based ICs and EALs are parameters or conditions that are measurable over some range using plant instrumentation (e.g., core temperature, reactor coolant level , radiological effluent , etc.). When one or more of these parameters or conditions are offnormal , reactor operators will implement procedures to identify the probable cause(s) and take corrective action. Fission product barrier-based I Cs and EALs are the subset of symptom-based EALs that refer specifically to the level of challenge to the principal barriers against the releas e of radioactive material from the reactor core to the environment.

These barriers are the fuel cladding , the reactor coolant system pressure boundary , and the containment.

The 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 I Cs and EALs define a variety of specific occurrences that have potential or actual safety significance.

These include the failure of an automatic reactor scram/trip to shut down the reactor , natural phenomena (e.g., an earthquake), or man-made hazards such as a toxic gas release. 3.3 NSSS-MNGP SPECIFIC DESIGN DIFFERENCES CONSIDERATIONS The NET 99 01 en:iergeAcy classificatioA scheme accouAts for the desigri differeAces betweeA PW Rs aAd BWRs by specifyiAg eAbs UAique to eaeh type of Nuelear Steam Supply System (N88S). There are also sigAificaAt desigA differeAees amoAg PWR }J"888s; therefore , guidanee is provided to aid iA the development of EALs appropriate to different

'PWR N8S8 types. Where Aecessary , development guidance also addresses unique eonsiderations for advaneed non passive reactor designs such as the Advanced BoiliAg Water Reaetor (ABWR), the Advanced Pressurized Water Reactor (APWR) and 11 the E*1olutionary Power Reactor (EPR). MNGP uses a single cyc l e , forced circu l at i on , l ow power de n s i ty boiling water reactor (BWR). Ge n era l Electr i c Compa n y d es i gned the p l a n t and supp li ed t h e nuc l ear steam s u pp l y system (NSSS), the i ni t i al reactor fue l , and t u r b i n e-generato r unit a n d i ts re l ated systems. Th i s design is identified as "BWR-3" by Ge n era l E l ectr i c. TCs and EAL thres h o l ds for a BWR NSSS have been appropr i ate l y in corporated i n to the MNGP emerge n cy c l assificat i on scheme. T h e reactor coo l ant syste m (a l so ca ll e d t h e reactor p r i mary system) i ncl u des t h e reactor vesse l; the 2-l oop reactor coo l ant rec i rcu l ation system with its pumps , pipes and va l ves; t h e m a i n steam pip i ng up to t h e main steam i solat i on valves; safety/relief va l ves; and the reactor aux ili ary systems p i pi n g. The reactor vesse l contains the reactor core and s u pporting structure , stea m separator a n d dryer assem bli es , jet pumps , control r od guide t ub es, and t h e R eactor Fee d wa t er , E m e r ge n cy Co r e Coo lin g Syste m (ECCS), a n d Sta nd by L i q ui d Control Syste m spargers.

T h e Pr i mary Co n ta i nment System, co n s i st in g of a stee l li ght-bu l b-shaped drywe ll , a stee l do u ghnut-sha p ed p r essu re s u ppression c h a m ber , and i n terconnecting vent pipes , provides t h e fi r st co n ta inm e n t bar ri e r s ur rou n d ing t h e r eactor vesse l and reactor prima r y system. T h e primary co n ta i nment system is des i gned to accom m odate t h e pressures , te m peratures, a n d hydrodynamic loads w hi ch would resu l t from , or occur subseq u ent to a post ul ated l oss-of-coo l a n t acc id ent (LOCA) w i th i n the p ri mary conta i nment and safety r e li ef va l ve o p eratio n s. A n y l eakage fro m t h e Pri m ary Containment System i s to t h e Seco n dary Co n ta inm ent Sys t e m , cons i st in g of the reactor b u i ldin g, t he p l ant S t a nd by Gas T r eat m ent Syste m , and the p l a n t ma in stack. T he p rim ary safeg u ards funct i o n s of the seco n dary co nt a i nment are to mi nimize ground level r e l ease of airborne radioactive mate ri a l s , a n d to prov i de for contro ll ed , fi l tered , e l evated release of secondary co n ta i nment at m osphere und er postu l ated design basis accident conditions.

Developers will need to consider the rele*t'ant aspects of their plant's design and operating eharaeteristies when eom'erting the generie guidance of tl'li s doeument into a site specific classification scheme. The goal is to maintain as much fidelity as possible to the intent of ge1'leric 1Gs and EA Ls within the constraints imposed by tl'le plant design and operating characteristics.

To this end , developers of a scheme for an advanced non passive reactor may need to add , modif)' or delete some information contained in this document; these changes will be reviewed for acceptability by the " NRG as part of the scheme approval process. The guidance in NEI 99 01 i s not applicable to advanced pa ss i*1e light water reactor designs. An Emergeney Classification Seheme for this type of plant should be developed in accordance with NE! 07 0 I , A1etheri8iegyfer Deo;eiepment

&jEmergen e y Actien Le*;eJ s, Adv e meed Pessio;e Light Weter Reecter s. 3.4 ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION The scheme's generic information is organized by Recognition Category in the following order.

A R -Abnormal Radiation Levels I Radiological Effluent -Section 6

C -Cold Shutdown I Refueling System Malfunction

-Section 7

E -Independent Spent Fuel Storage Installation (ISFSI) -Section 8 12

F -Fission Product Barrier -Section 9

H -Hazards and Other Conditions Affecting Plant Safety -Section 10

S -System Malfunction

-Section 11

PD Permanently Defueled Station Appendix C Each Recognition Category section contains a matrix showing the ICs and their associated emergency classification levels. The following information and guidance is provided for each IC:

ECL -the assigned emergency classification level for the IC.

Initiating Condition

-provides a summary description of the emergency event or condition.

Operating Mode Applicability

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

Emergency Action Level(s) -Provides reports and indications that are considered to meet the intent of the IC. Developers should address each example EAL. If the generic approach to the development of an example EAL cannot be used (e.g., an assumed instrumentation range is not available at the plant), the de*,reloper shm.1ld attempt to specify an alternate means for identifying entry into the IC. For Recognition Category F , the fission product barrier thresholds are presented in a table s applicable to BWRs and P\\£.Rs MNGP , and arranged by fission product barrier and the degree of barrier challenge (i.e., potential loss or loss). This presentation method shows the synergism among the thresholds , and supports accurate assessments.

Basis -Provides background information that explains the intent and application of the IC and EALs. In some cases , the basis also includes relevant source information and references.

Notes Information that supports the development of the site specific ICs and EALs. This may include clarifications, references , e>tamples, instructions for calculations, etc. Developer notes should not be included in the site's emergency classification scheme basis document.

DeYelopers may elect to include information resulting from a developer note action in a basis section.

ECL Assignment AttFibutes Located within the DeYeloper Notes section, specifies the attribute ttsed for assigning the IC to a given HCL. 3.5 IC AND EAL MODE APPLICABILITY The NEI 99-01 emergency classification scheme adopted by MNGP was developed recognizing that the applicability of I Cs and EALs will vary with plant mode. For example, some symptom-based ICs and EALs can be assessed only during the power operations , startup , or hot standby/shutdown modes of operation when all fission product 13

barriers are in place, and plant instrumentation and safety systems are fully operational.

In the cold shutdown and refueling modes , different symptom-based ICs and EALs will come into play to reflect the opening of systems for routine maintenance , the unavailability of some safety system components and the use of alternate instrumentation.

The following table shows which Recognition Categories are applicable in each plant mode. The ICs and EALs for a given Recognition Category are applicable in the indicated modes. MODE APPLICABILITY MATRIX FOR MNGP Category Mode AR c E F H s Power Operations x x x x x Startup x x x x x Hat x --x x x Hot Shutdown x x x x x Cold Shutdown x x x x Refueling x x x x Defueled x x x x Permanent l y x 9efue l ea 14 x Mode I 2 3 4 5 MNGP Operating Modes Defueled (None): Title Power Operation Startup Hot Shutdown (a) Cold Shutdown (a) Refueling Cbl Run Po*t'l'er Operations (I): Mode 8witeh in Run 8tartup (2): Mode 8witch in 8tartup/l=fot 8tandby or Refuel (with all vessel head bolts fully tensioned)

I lot 8hutdo'tvn (3): Mode 8witeh in 8hutdown , Average Reaetor Coolant Temperature

>200 °F Cold 8hutdown (4): Mode 8witeh in 8hutdown , Average Reaetor Coolant Temperature<

200 °F Refueling (5): Mode 8v1itch in 8hutdown or Refuel , &Rd one or more vessel head bolts less than full;' tensioned.

All fuel removed from the reaetor vessel (i.e., full core offload during refueling or e J (tended outage). Reactor Mode Average Reactor Switch Po s ition Coolant Te mperatur e (°F) NA Refuel(a) or Startup/Hot Standby NA Shutdown > 212 Shutdown Shutdown or Refuel NA Defueled (None): All fuel removed from the reactor vessel (i.e., full core offload durin g refueling or extended outage). (a) All reactor vessel head closure bolts fully tensioned. (b) One or more reactor vessel head closure bolts less than fully ten s ioned. 15 4 SITE SPECIFIC MNGP SCHEME DEVELOPMENT GUIDANCE This section provides detailed guidance for developing a site specific emergenC)' classification scheme. Conceptually, the approach discussed

.. liere mirrors the approach used to prepare emergency opera-ting procedures generic material prepared by reactor vendor owners groups is converted by each nuclear power plant into site specific emergency operating procedures.

Likewise, the emergency classification scheme developer will use the generic guidance in NEl 99 01 to prepare a site specific emergency classification scheme and the associated basis document.

lt is important that the NEI 99 01 emergency classification scheme be implemented as an integrated package.

Selected use of portions of this guidance is strongly discouraged as it will lead to an inconsistent or incomplete emergency classification scheme that will likely not receive the necessary regulatory approval.

4.1 GENERAL IMPLEMENTATION GUIDANCE The guidance in NEl 99 01 is not intended to be applied to plants " as is"; however, deve l opers should attempt to keep their site specific schemes as close to the generic guidance as possib l e. The goal is to meet the intent of the generic Initiating Conditions (JCs) and Emergency Action Levels (EALs) *.vithin the eonte)H of site specific characteristics locale , plant design , operating features , terminology , etc. Meeting this goal will result in a shorter and less cumbersome NRG review and approval process , c l oser alignment

'Nith the schemes of other nuclear power plant sites and better positioning to adopt future industry wide scheme enhancements.

When properly developed, the MNGP ICs and EALs should were developed to be unambiguous and readily assessable.

As discussed in Section 3 , the generic guidance includes ICs and e>rnmple BALs. It is the intent of this guidance that QQ!h be included in site specific documents as each serves a specific purpose. The IC i s 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 MNGP ICs and EALs were developed with input from key stakeholders such as Operations , Training , Radiation Protection, Chemistry, and Engineering.

MNGP specific indications , parameters , and values are consistent with licensing basis documents , plant procedures , training , calculations , and drawin g s. If some feature of the plant location or design is not compatible

'With a generic IC or EAL , efforts should be made to identify an alternate IC or BAL. If an re or BAL includes an exp I icit reference to a mode dependent technical specification li mit that is not applicable to the plant , then that JG and/or EAL need not be included in the site specific scheme. In these eases , developer s must provide adequate documentation to justify why the IC and/or EAL '+Vere not incorporated (i.e., sufficient detail to allow a third party to understand the decision not to incorporate the generic guidance).

Useful acronyms and abbreviations associated with the N E I 99 01 MNGP emergency classification scheme are presented in Appendix A , Acronyms and Abbreviat ion s. -Si-te-16 specific entries may be added if necessary Those specific to MNGP were included to be consistent with site terminology, s ite procedures , and training. Many words or terms used in the }lei 99 01 MNGP emergency classification sc h eme have scheme-specific definitions.

T h ese words and terms are id ent ifi ed by being set in all capital l etters (i.e., ALL CAPS). The definitions are presented in Appendix B , Definitions.

Below are examples of acceptable modifications to the generic guidance.

These may be incorporated depending upon site developer and user preferences.

The lCs within a Recognition Category may be placed in reverse order for presentation purposes (e.g., start with a General Emergency at the left/top of a user aid , follo*Ned by Site Area Emergency , Alert and }J"OUE).

The Initiating Condition numbering may be changed.

The first letter of a Recognition Category designation may be changed, as follows , provided the change is carried through for all of the associated IC identifiers.

R may be used in lieu of A

M may be used in lieu of 8 for ex.ample, the Abnormal Radiation Levels I Radiological Effluent category designator "A" (for Abnormal) may be changed to "R" (for Radiation).

This means that the associated ICs 'Nould be changed to RU I , RU2 , R/\, 1 , etc.

The ICs and EAbs from Recognition Categories 8 and C may be incorporated into a common 19resentation method (e.g., one table) provided that all related notes and mode applicability reEtuirements are 1'l'laintained.

The ICs and EAbs for Emergency Director judgment and security related events may be placed under separate Recognition Categories.

The terms EAL and threshold ma)' be used interchangeably. The material in the Develo19er 11-Jotes section is included to assist deYelopers

.vith crafting correct IC and EAL statements.

This material is not reEtuired to be in the final emergency classification scheme basis document.

4.2 CRITICAL CHARACTERISTICS As discussed above , developers are encouraged to keep their site s19ecific schemes as close to the generic guidance as possible.

When crafting the scheme , de,,*elopers should satisfy themselves MNGP ensured that certain critical characteristics have been were met. These critica l characteristics are li sted below.

The ICs , EALs, Operati n g Mode Appl i cability criteria , Notes and Basis information are consistent with ind u stry guidance; while the actual word in g may be different from NEI 99-0 I Revision 6 , the classification intent is maintained.

With respect to Recognition Category F , a site speeific the MNGP scheme fRHSt-include s a some type e:f:.user-aid to facilitate timely and accurate classification of fission product barrier 17 losses and/or potential losses. The user-aid logic must be is consistent with the classification logic presented in Section 9.

The ICs, EALs, Operating Mode Applicability criteria, Notes and Basis information are technically complete and accurate (i.e., they contain the information necessar y to make a correct classification).

EAL statements use objective criteria and observable values.

ICs , EALs, Operating Mode Applicability and Note statements and formatting consider human factors and are user-friendly.

The sc heme facilitates upgradin g and dovmgrading of the emergency classification where necessar y.

The scheme facilitates classification of multiple concurrent events or conditions.

4.3 INSTRUMENTATION USED FOR EALS Instrumentation referenced in EAL statements should include that described in the emergency plan section 'Nhich addresses l 0 CPR 50.4 7(b)(8) and (9) and/or Chapter 7 of the FSAR. Instrumentation used for EALs need not be safet)' related , addressed by a Technical Specification or ODCM/RETS control requirement , nor powered from an emergency pO'Ner source; however , EAL developers should strive to incorporate MNGP EAL thresholds utilize instrumentation that i s reliable and routinely maintained in accordance with site programs and procedures. Alarms referenced in EAL statements should be are those that are the most operationally significant for the de scri bed event or condition.

Scheme de*velopers should MNGP personnel have ensure d that specified values used as EAL setpoints are within the calibrated range of the referenced in st rumentation , and consider any automatic instrumentation function s that ma y impact accurate EAL assessment.

In addition, EAL setpoint values should do not use terms such as "off-scale low" or "off-scale high" since that type of reading may not be readily differentiated from an instrument failure. Findings and violations related to EAL instrumentation issues may be located on the 1'JRC website lf instrumentation failures occur that have EALs associated with them (e.g., process radiation monitors) compensatory means of implementation may be used as described in plant procedures. 4.4 PRESENTATION OF SCHEME INFORMATION TO USERS The US Nuclear Regulatory Commission (NRC) expects licensees to establish and maintain the capability to assess, classify and declare an emergency condition promptl y within 15 minutes after the availability of indications to plant operators that an emergency action level has been , or may be , exceeded.

\Vhen 'Nriting an emergenC)' classification procedure and creating related user aids , the developer must determine the presentation method(s) that best supports the end users by facilitating accurate and timely emergency classifieation The MNGP emergency classification procedure and user aid (EAL Matrix) have been developed to facilitate accurate and timely classification. To this end, developers should consider the following point s have been considered.

The first users of an emergency classification procedure are the operators in the Control Room. During the allowable classification time period, they may have 18 responsibility to perform other critical tasks , and will likely have minimal assistance in making a classification assessment.

As an emergency situation evolves , members of the Control Room staff are likely to be the first personnel to notice a change in plant conditions.

They can assess the changed conditions and , when warranted , recommend a different emergency classification level to the Technical Support Center (TSC) and/or Emergency Operations Facility (EOF).

Emergency Directors in the TSC and/or EOF will have more opportunity to focus on making an emergency classification , and will probably have advisors from Operations available to help them. Emergency The MNGP emergency classification scheme information for end users should be is presented in a manner with which licensed operators are most comfortable.

Developers will need to work closely with representatives Input from the Operations and Operations Training Departments has. been used to assist in the develop ment of readily usable and easily understood classification tools (e.g., a procedure and related user atds EAL Matrix). Tf necessary , an alternate method for presenting emergency classification scheme information may be developed for use by Emergency Directors and/or Offsite Response Organization personnel.

A-The MNGP 'ncallboard EAL Matrix is an acceptable presentation method provided that #-contains all the information necessary to make a correct emergency classification. +hts MNGP wallboard EAL Matrix information includes the ICs , Operating Mode Applicability criteria , EALs and Notes. Notes may be kept with each applicable EAL or moved to a common area and referenced; a reference to a }fote is acceptable as long as the information is are adequately captured on the wallboard EAL Matrix and pointed to by each applicable EAL;. Basis information nee&-is not be-included on a-the MNGP wallboard EAL Matrix but it should be is readily available to emergency classification decision-makers.

In some cases , it may be advantageous to MNGP has develop ed two wallboards matrices -one for use during power operations , startup and hot conditions, and another for cold shutdown and refueling conditions.

Alternative presentation methods for the Recognition Category F ICs and fission product barrier thresholds are acceptable and include flow charts, block diagrams , and checklist type tables.

Developers must ensure that the site specific method addresses all possible threshold combinations and classification outcomes shovm in the BWR or PWR EAL fission product barrier tables. The NRG staff considers the presentation method of the Recognition Category F information to be an important user aid and may request a change to a particular proposed method if, among other reasons, the change is necessary to promote consistency across the industry.

a1313r013riate , ti-le Jl-Jates sl-iawA iA ti-le geAerie g1:1icfoAee ty13ieally iAelude ti-le e\'eAtfeaAditiaA BCL aAd the d1:1rati0A time s13eeified iA ti-le EAL. lfdevelapers prefer ta ha\'e several !Cs refereAee a eammaA NOTE; aA a *n*alleeard display , it is aeeeptaele ta remave ti-le EGL aAd time eriterieA aAd use a geAerie statemeAt.

far example , a e011'lm0A

'NOTE eauld read "Tl-le EmergeAey Direeter sh01:1ld deelare the premptly upeA deter11'liAiAg that ti-le applieable EAL time has eeeA e)rneeded, er will likely be 0)£6eeded." 19 4.5 INTEGRATION OF ICs/EALs WITH PLANT PROCEDURES A rigorous integration of IC and EAL references into plant operating procedures is not recommended.

This approach would greatly increase the administrative controls and workload for maintaining procedures.

On the other hand , performance challenges may occur if recognition of meeting an IC or EAL is based solely on the memory of a licensed operator or an Emergency Director , especially during periods of high stress. Developers should consider placing appropriate visual cues (e.g., a step , note , caution , etc.) in plant procedures alerting the reade_r/user to consult the site emergency classification procedure.

Visual cues (e.g., a ste p , n o t e, ca u t i o n , etc.)could be placed are includ e d in pl a nt p roce dur es (inc l udin g emergency operating procedures , abnormal operating procedures , alarm response procedures , and normal operating procedures

), as appropri ate, a l e rtin g th e r ea d e r/u ser t o c o n s ult the s ite e m e r ge ncy c l ass ifi ca ti o n proc edu r e. that apply to cold shutdo'Nn and refueling modes. As an e>1.ample , a step , note or caution could be placed at the beginning of an RCS leak abnormal operating procedure that reminds the reader that an emergency classification assessment should be performed.

4.6 BASIS DOCUMENT A basis document is an integral part of an emergency classification scheme. The material in this document supports proper emergency classification decision-making by providing informing background and development information in a readily accessible format. It can be referred to in training situations and when making an actual emergency classification , if necessary.

The document is also useful for establishing configuration management controls for EP-related equipment and explaining an emergency classification to offsite authorities.

The content of the MNG P basis document sheu-ki include s , at a minimum , the following:

An site specific MNG P Mode Applicability Matrix and description of operating modes , similar to that presented in section (Sect i o n 3.5).

A discussion of the emergency classification and declaration process reflecting the material presented in (Section 5). This material may be edited as needed to align with site specific em,ergency plan and implementing procedure requirements.

Each Initiating Condition along with the associated EALs or fission product barrier thresholds , Operating Mode Applicability , otes and Basis information.

A listing of acronyms and defined terms , similar to that presented in Appendices A and B , respectively. This material may be edited as needed to align with site-specific characteristics.

Any site specific background or technical appendices that the developers believe *.vould be useful in eKplaining or using elements of the emergency classification scheme. A-Th e MN GP Basis section should d oes not contain information that could modify the meaning or intent of the associated IC or EAL. Such information should be incorporated within the IC or BAL statements, or as an eAL Jl-Jote. Information in the Basis should oo-1-y i s u sed o nl y to clarify and inform decision-making for an emergenc y classification.

Basis information should be i s readily available to be referenced , if necessary , by the 20 Emergency Director.

for example , a A copy of the MNGP basis document could be is maintained in the appropriate emergency response facilities.

Because the information in a basis document can affect emergency classification decision-making (e.g., the Emergency Director refers to it during an event), the " NRG staff expects that changes to the MNGP basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q). 4. 7 EAL/THRESHOLD REFERENCES TO MNGP AOP AND EOP SETPOINTS/CRITERIA As reflected in the generic guidance , tT he criteria/values used in several EALs and fission product barrier thresholds may be were drawn derived from a plant MNGP's AOPs and EOPs. This approach is intended to maintain good alignment between operational diagnoses and emergency classification assessments.

Developers should MNGP has ¥ef-i.fy-verified that appropriate administrative controls are in place to ensure that a subsequent change to an AOP or EOP is screened to determine if an evaluation pursuant to 10 CFR 50.54(q) is required.

4.8 DEVELOPER ANO USSR FSSOBACK Questions or comments concerning the material in this document may be directed to the l>JEl Emergency Preparedness staff , NEI EAL task force members or submitted to the Emergency Preparedness frequently Asked Questions process. 21 5 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 5.1 GENERAL CONSIDERATIONS When making an emerge n cy c l assificat i on , the Emergency Director must consider all information having a bear in g on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability , Notes and the informing Basis information.

In the Recognition Category F matrices , EALs are referred to as Fis s ion Product Barrier Thresholds; the thresholds serve the same function as an EAL. NRC regulations require the licen see to establish and maintain the capability to assess , classify , and declare an emergency cond iti on within 15 minutes after the ava ilabilit y of indications to plant operators that an emergency act i on l evel has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency c la ssification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG

-01 , Interim Staff Guidance , Emergency Planning for Nuclear Power Plants. All emergency classification assessments should be based upon valid indications , reports or conditions. A valid indication , report , or condition, is one that has been verified through appropriate means such that there i s no doubt regarding the indicator's operability , the condition's existence, or the report's accuracy.

For example, validation could be accomplished through an instrum ent channel check , response on related or redundant indicators , or direct observat i on by plant personnel.

The va lid ation of indications should be completed in a manner that supports timely emergency declaration.

For ICs and EALs that have a stipulated time duration (e.g., 15 minutes , 30 minutes , etc.), the Emergency Director sho uld not wait until the app licabl e time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded , or will likely exceed , the applicab l e time. If an ongoing radio l ogical release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, abse nt data to the co ntrar y. A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that 1) the activity proceeds as planned and 2) the plant remains within the limits imposed by the operating lic ense. 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 a ll aspects of the operating license provided that the activity proceeds and concl ud es as expected.

Events or conditions of this type may be subject to the reporting requirements of 10 § CFR 50.72. The assessment of some EALs i s based on the results of analyses that are necessary to ascertain whether a spec ific EAL threshold has been exceeded (e.g., dose assessments , chemistry samp lin g , RCS leak rate calculation , etc.); the EAL and/or the associated basis discussion w ill identify the necessary analys i s. In these cases , the 15-minute declaration period starts w ith the ava ilabilit y of the ana l ysis results that show the threshold to be 22 exceeded (i.e., this is the time that the EAL information is first available).

The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).

While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification , a provision for classification based on operator/management experience and judgment is still necessary.

The NEI 99 1-MNGP scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated into the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier. 5.2 CLASSIFICATION METHODOLOGY To make an emergency classification , the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded.

The evaluation of an EAL(s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded , then the IC is considered met and the associated ECL is declared in accordance with plant procedures.

When assessing an EAL that specifies a time duration for the off-normal condition , the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock." For a full discussion of this timing requirement, refer to ISG-01. 5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS When multiple emergency events or conditions are present , the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared.

For example:

If an Alert EAL and a Site Area Emergency EAL are met , whether at one unit or at two different units , a Site Area Emergency should be declared.

There is no "additive" effect from multiple EALs meeting the same ECL. For example:

If two Alert EALs are met , whether at one unit or at ti.vo different units , an Alert should be declared.

Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02 , Clarification of N RC Guidance for Emergency Notifications During Quickly Changing Events. 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION The mode in effect at the time that an event or condition occurred , and prior to any plant 23 or operator response, is the mode that determines whether or not an IC is applicable.

If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level i s still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared).

Once a different mode is reached, any new event or condition , not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.

For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes , even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response.

In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 5.5 CLASSIFICATION OF IMMINENT CONDITIONS Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meetin g or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT).

If , in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as ifthe EAL has been met. While applicable to all emergency classification levels , this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.

5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING Once a classification level is declared , no downgrade to a lower classification will be allowed. The MNGP E mergency Plan and classification E PIPs provide the applicable guidance for transition to Termination and/or Recovery. An EGL may be dovmgraded

\Vhen the event or condition that meets the highest IC and EAL no longer e><ists , and other site specific downgrading requirements are met. If downgrading the EGL is deemed appropriate , the ne'N EGL 'NOuld then be based on a lo*.ver applicable lG(s) and EAL(s). The EGL may a l so simply be terminated.

The following approach to downgrading or terminating an EGL is recommended.

E(;l, Aetian When (;0editi0n Na bangel' Elitists Um1sual Event Terminate the emergency in accordance "vith plant . Alert .Qowngrade or terminate the in accordance with plant 13rocedures.

Site Area Emergency Qowngrade or terminate the emergency in with no long term 13lant accordance "vith 13lant procedures.

damage 24 8ite Afea emeFgenoy

+eFminate the emeFgenoy and entef with long tenn plant reoovefy in accofdance 1 Nith plant Elamage 13rnoeElt:1Fes.

Genernl emeFgenoy

+eFminate the anEI entef Feoovefy in accoFElance with plant f3FOCeet:tfeS.

As noted above, g G uidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02. 5.7 CLASSIFICATION OF SHORT-LIVED EVENTS As discussed in Section 3.2, event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus , over before the emergency classification assessment can be completed.

If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration.

Examples of such events include an earthquake or a failure of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip Of an earthquake. 5.8 CLASSIFICATION OF TRANSIENT CONDITIONS Many of the I Cs and/or EALs contained in this document employ time-based criteria.

These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted.

In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes).

The following guidance should be applied to the classification of these conditions. EAL momentarily met during expected plant response -In instances where an EAL is briefly met during an expected (normal) plant response , an emergency declaration is not warranted provided that associated systems and components are operating as expected , and operator actions are performed in accordance with procedures.

EAL momentarily met but the condition is corrected prior to an emergency declaration

-If an operator takes prompt manual action to address a condition , and the action is successful in correcting the condition prior to the emergency declaration , then the applicable EAL is not considered met and the associated emergency declaration is not required.

For illustrative purposes , consider the following example. An A TWS occurs and the amtplosive mi>tture) exists inside containment

UNPLANNED inerease rise in containment pressure of greater than 1.84 psig
Two or more Reactor Building areas exceed Max Safe Radiation Levels (C.5-1300, Table X)Seeondary eontainment radiation monitor reading above (site speeifie value) [BWR]
If CONTATNMBNT CLOSURE SECONDARY CONTAINMENT is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.
46 Basis: SECONDARY CONTAINMENT:
SECONDARY CONTAINM E NT includes the Reactor Building (including the HPCT Building), the Standby Gas Treatment System , the Offgas Dilution Fans , and connecting pipes and ducts. SECONDARY CONTAINMENT is isolated along with an automatic initiation of the Standby Gas Treatment System to minimize radiological releases to the environment.
UNPLANNED:
A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses the inabilit y to restore and maintain reactor vessel level above the top of active fuel with containment challenged.
This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup , decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored , fuel damage is probable.
With CONTAINMB+/-'ff CLOSURB SECONDAR Y CONTAINMENT not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment.
If CONTAfl'J"MENT CLOSURB SECONDARY CONTAINMENT is re-established prior to exceeding the 30-minute time limit , then declaration of a General Emergency is not required.
The existence of an explosive mixture means , at a minimum , that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity.
It therefore represents a challenge to Containment integrity.
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment.
If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage , it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service , operators may use the other listed indications to assess whether or not containment is challenged.
In EAL CGl .2.b , the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor , assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).
It also allows sufficient time for performance of actions to terminate leakage , recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RPV level may be caused by instrumentation and/or power failures , or water level dropping below the range of available instrumentation. If water level cannot be monitored , operators may determine that an inventory loss is occurring by observing changes in 47 sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. These EALs address concerns raised by Generic Letter 88-17 , Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449 , Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06 , Guidelines for Industry Actions to Assess Shutdown Managem e nt. MNGP Basis Reference(s):
1. Ops Man C.1 (STARTUP PROCEDURE)
2. Ops Man C.4-B-04.01.F (LEAK INSIDE PRIMARY CONTAfNMENT)
3. Ops Man C.6-004-B-13 (DRYWELL EQUIP DRAIN LEAK RATE HI) 4. Ops Man C.6-004-B-l 7 (DRYWELL FLOOR DRAIN SUMP HT) 5. Ops Man C.6-004-8-18 (DRYWELL EQUIP DRAfN LEAK RA TE CHANGE HT) 6. Ops Man C.6-065-A-06 (REFUELING:
HIGH LEAKAGE) 7. Emergency Operating Procedure C.5.1-1100 (RPV CONTROL) 8. Emergency Operating Procedure C.5.1-1300 (SECONDARY CONTAINMENT CONTROL) 9. USAR Section 5.2, Containment System -Primary Containment System 10. USAR Section 7.9-1 , Plant Instrumentation and Control Systems-Accident Monitoring Instrumentation
11. MNGP Calculation CA-04-202 , Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level 48 CS1 ECL: Site Area Emergency Initiating Condition:
Loss of RPV inventory affecting core deca y heat remo val capability.
Operating Mode Applicability:
Cold Shutdown, Refueling Emerge n cy Action Levels: (CS 1.1 or CS 1.2 or CS 1.3) Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded , or will likel y be exceeded. CSl.1 a. CONTAINMENT CLOSUR5 SECONDARY CONTAINMENT not established.
CSl.2 CSl.3 Basis: AND b. RPV level les s than--4 7 in.(site specific level). a. AND b. a. AND b.
CLOSURE SECONDARY CONTAINMENT established.
RPV level le ss than -126 in. (TAF).(site specific le,*el). RPV l evel cannot be monitored for 30 minutes or longer. Core uncovery is indicated by AN¥-EITHER of the follow i ng:
fR efueling Floor Site specific radiation monitor} reading greater than 3 R/hr (site specific value)
UNPLANNED increase rise in drywell floor or equipment drain sumpspecific sump and,Lor tank) l evels of s ufficient magnitude to indicate core uncovery * (Other site specific indications)
SECONDARY CONTAINMENT:
SECONDARY CONTAINMENT includes the Reactor Building (including the HPCI Building), the Standby Gas Treatment System , the Off gas Dilution Fans , and connecting pipes and ducts. SECONDARY CONTAINMENT is isolated along with an automatic initiation of the Standby Gas Treatment System to minimize radiological releases to the environment.
UNPLANNED:
A parameter change or an event that is not I) the re s ult of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses a significant and prolonged loss of RPY i nventor y contro l and makeup capabi li ty l eading to IMMINENT fue l damage. The lost inventory may be due to a RCS component failure, a l oss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of p l ant functions needed for protection of the public and thu s warrant a Site Area Emergency declaration.
Following an extended loss of core decay heat removal and inventory makeup , decay heat will cause reactor coolant boilin g and a further reduction in reactor vessel leve l. If RCS/reactor 49 vessel level cannot be restored , fuel damage is probable.
Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE SECONDARY CONTAINMENT following a loss of heat removal or RCS inventory control functions.
The difference in the specified RCS/reactor vessel levels of EALs CSl.1.b and CSl.2.b reflect the fact that with CONTAINMENT CLOSURE SECONDARY CONTAINMENT established , there is a lower probability of a fission product release to the environment.
In EAL CSl.3.a , the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess a nd correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).
It also allows sufficient time for performance of actions to terminate leakage , recover inventory control/makeup equipment and/or restore level monitoring.
The inability to monitor RPV level may be caused b y instrumentation and/or power failures, or water level dropping below the range of available instrumentation.
If water level cannot be monitored , operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. These EALs address concerns raised by Generic Letter 88-17 , Loss of Decay Heat Removal; SECY 91-283 , Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449 , Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industr y Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CG 1 or RG 1. MNGP Basis Reference(s):
1. Ops Man B.01.01-03 (REACTOR AND VESSEL ASSEMBLY)
2. Ops Man B.05.06 (PLANT PROTECTION SYSTEM) 3. Ops Man C.1 (STARTUP PROCEDURE)
4. Ops Man C.4-B.04.0 I .F (LEAK INSIDE PRIMARY CONTAINMENT)
5. Ops Man C.6-003-A-38 (REACTOR LOW LOW LEVEL) 6. Ops Man C.6-004-B-13 (DRYWELL EQUTP DRAIN LEVEL RA TE HT) 7. Ops Man C.6-004-B-17 (DRYWELL FLOOR DRAIN SUMP HJ LEVEL) 8. Ops Man C.6-004-B-18 (DRYWELL EQUlP DRAIN LEAK RATE CHANGE HI) 9. Tech Spec Table 3.3.5.1-1 (EMERGENCY CORE COOLING INSTRUMENTATION)
10. Tech Spec 3.4.4 (RCS OPERATIONAL LEAKAGE) 11. NX-7831-197-1 , Reactor Vessel and Internals 50
12. MNGP Calculation CA-95-074 , Low Low Reactor Water Level Group 1 and 3 Conta i nment Isolat i on Setpoint Calcu l ation 13. MNGP Ca l culation CA-04-202, Dose Rates to CHRRM Detector s Due to Drop in RPV Water Level 51 CA1 ECL: Alert Initiating Condition:
Loss of RPV inventory.
Operating Mode Applicability:
Cold Shutdown , Refueling Emergency Action Levels: (CA 1.1 or CA 1.2) Note: The Emergency Director should declare the Alert promptl y upon determining that 15 minutes has been exceeded , or will likely be exceeded. CAl.1 Loss of RPV inventory as indicated by l evel less than -47 in.(site specific level Low Low E:CC8 actuation setpoint/Level 2). CAl.2 a. RPV level cannot be monitored for 15 minutes or longer Basis: AND b. UNPLANNED increase rise in drywell floor or equipment drain sumspecific sump and/or tank) levels due to a loss of RPV inventory.
UNPLANNED:
A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier).
This condition represents a potential substantial reduction in the level of plant safety. For EAL #CA 1.1 , a lo wer in g of water level below -4 7 in. (site specific level) (Low-Low ECCS actuation setpoint) indicates that operator actions have not been successfu l in restoring and maintaining RPV water l evel. The heat-up rate of the coolant will increase as the availab l e water inventory is reduced. A continuing decrease in water level wi ll lead to core uncovery.
Although related , EAL #CA 1.1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Heat Removal suct i on point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3. For EAL #CA 1.2 , the inability to monitor RPV l evel may be caused by instrumentation and/or power failures , or water level dropping below the range of available instrumentation. If water le vel cannot be monitored , operators may determine that an inventory l oss i s occurr in g by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS 1 52 If the RPV inventory level continues to lower , then escalation to Site Area Emergency would be via IC CSL MNGP Basis Reference(s)
l. Ops Man B.Ol.0 1-03 (REACTOR AND VESSEL ASSEMBLY -INSTRUMENTATION AND CONTROLS)
2. Ops Man B.05.06 (PLANT PROTECTION SYSTEM) 3. Ops Man C.4-B.04.01.F (LEAK INSIDE PR I MARY CONTAINMENT)
4. Ops Ma n C.6-003-A-38 (REACTOR LOW LOW LEVEL) 5. Ops Man C.6-004-8-13 (D R YWELL EQU I P DRAIN LEAK RATE HI) 6. Ops Man C.6-004-B-17 (DRYWELL FLOO R DRAIN SUMP HI) 7. Ops Man C.6-004-B-18 (DR YWELL EQU T P ORA JN LEAK RA TE CHAN GE HI) 8. 9040 (TEMPORARY VESSEL LEVEL INSTRUMENTATION INSTALLATION AND RESTORATION)
9. Tec h Spec Tab le 3.3.5. l-1 (EMERGENCY CORE COOLING SYSTEM INSTRUMENTATION)
10. Ops Man C.6-065-A-06 (REFUELING HIGH LEAKAGE) 53 CA2 ECL: Alert Initiating Condition:
Loss of all off site and all onsite AC power to emergency essential buses for 15 minutes or longer. Operating Mode Applicability:
Cold Shutdown , Refueling , Defueled Emergency Action Levels: Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been e x ceeded, or will likely be exceeded.
CA2. l Loss of ALL offsite and ALL onsite AC Power to essential buses 15 and l 6 (site specific emergency buses) for 15 minutes or longer. Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling , containment heat removal/pressure control , spent fuel heat removal and the ultimate heat sink. This EAL is indicated by the loss of all offsite and onsite AC power to the 4 l60V essential buses. Onsite sources include 11 and 12 Emergency Diesel Generators.
Offsite resources include 2R, lR , and lAR Transformers.
If power is available from these sources , but is not supplied to the 4160V buses for whatever reason, the condition is still considered a loss of power. When in the cold shutdown , refueling , or defueled mode , this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency essential bus to service. Additional time is available due to the reduced core decay heat load , and the lower temperatures and pressures in various plant systems. Thus , when in these modes , this condition represents an actual or potential substantial degradation of the level of safety of the plant. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC CS 1 or RS 1. MNGP Basis Reference(s):
l. USAR Section 8.2 , Plant Electrical Systems -Transmission System 2. USAR Section 8.3 , Plant Electrical Systems -Auxiliary Power System 3. USAR Section 8.4 , Plant Electrical Systems -Plant Standby Diesel Generator System 4. USAR Figure 8.4-1, Diesel Generation System One Line Diagram 5. NF-36175 , Single Line Diagram -Station Connection 54
6. Ops Man B.09.06-01 (4.16 KV STATION AUXILIARY
-FUNCTION AND GENERAL DESCRIPTION OF SYSTEM) 7. Ops Man C.4-B.09.02.A (STATION BLACKOUT)
8. Ops Man C.4-B.09.02.B (LOSS OF NORMAL OFF-SITE POWER) 9. Ops Man C.4-B.09.06.C (LOSS OF BUS 15 OR BUS 16) 10. Tec h Spec 3.8.2 (AC SOURCES -SHUTDOWN) 55 CA3 ECL: Alert Initiating Condition:
Inabilit y to maintain the plant in co l d s hutd ow n. Operating Mode Applicability:
Cold Shutdown, Refuel i n g Emergency Action Levels: (CA3.1 or CA3.2) Note: The Emergency Director should declare the Alert promptl y upon determining that the applicable time has been exceeded, or will likel y be exceeded. CA3.l UNPLANNED increase rise in RCS temperature to greater than (site specific Technical Specification cold shutdown temperature limit)212 °F for grea ter than the duration specified in the following table Table C2. Table C2: RCS Heat-up Duration Thresholds RCS Centainment Heat-up Duration Clesure SECONDARY CONTAINMENT Not intact Not Established 0 minutes Established 20 minutes* Intact N I A 60 minutes*
If an RCS heat remova l system i s in operation within this time frame and RCS temperature is bein g reduced, the EAL i s not applicable.
CA3.2 UNPLANNED RCS pressure increase rise greater than 10 psig (site specific pressure reading). Basis: SECONDARY CONTAINMENT:
SECONDARY CONTAINMENT includes the Reactor Building (including the HPCI Building), the Standby Gas Treatment System , the Offgas Dilution Fans, and connecting pipes and ducts. SECONDARY CONTAINMENT is isolated along with an automatic initiation of the Standby Gas Treatment System to minimize radiological releases to the environment.
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses conditions involving a loss of decay heat remo va l capabil i ty or a n addition of heat to the RCS in excess of that which can current l y be removed. Either cond i tion represent s an actua l or potential substantial degradation of t h e l evel of safety of the plant. A momentar y UNPLANNED excursion above the Technical Specification cold s hutdown temperature limit when the heat removal function is availab l e does not warrant a classification.
56 The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT GLOSURE SECONDARY CONTAINMENT is established but the RCS is not intac t. The 20-minute criterion was included to allow time for operator action to address the temperature increa se. The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of GONTl\JNMENT GLOSURE SECONDARY CONTAINMENT is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety. Finally, in the case where there is an increase in RCS temperature, the RCS is not intact and GONTAJNr..ffiNT GLOSURE SECONDARY CONTAINMENT is not established , no heat-up duration is allowed (i.e., 0 minutes).
This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment , and 2) there is reduced reactor coolant inventory above the top of irradiated fuel. EAL #CA3.2 provides a pres s ure-based indication of RCS heat-up. Escalation of the emergency classification level would be via IC CS 1 or RS 1. MNGP Basis Reference(s)
1. Ops Man B.01.0 1-03 (REACTOR AND VESSEL ASSEMBLY -INSTRUMENTATION AND CONTROLS)
2. Ops Man C.3 (SHUTDOWN PROCEDURE)
3. MNGP Emergency Plan Tab l e 1 3 , Instruments Available For Monitoring Major Systems 4. USAR Section 7.4 , P l ant Instrumentation and Control Systems -Reactor Vessel lnstrumentation
5. Tech Spec Tab l e l.1-1 (MODES) 6. Tech Spec 3.6.4.1 (SECONDARY CONTArNMENT)
7. Tech Spec 3.6.4.2 (SECONDARY CONTAINMENT ISOLATION VALVES (SC!Vs)) 8. Tech Spec 3.6.4.3 (STANDBY GAS TREATMENT (SGT) SYSTEM) 57 CA6 ECL: Alert Initiating Condition:
Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability:
Cold Shutdown , Refueling Emergency Action Levels: CA6.l Basis: a. The occurrence of ANY of the fo ll owing hazardous events:
Seismic event (Earthquake)
Internal or external flooding event
High winds or tornado strike
FIRE
EXPLOSION
River level greater than 919 ft el.
River level less than 900.5 ft el. ._ (s i te specific haz:ards)
Other event s with similar hazard characteristics as determined by the Shift Manager AND b. EITHER of the fo llo wing:
Event damage has caused indic ations of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.
The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode. EXPLOSION:
A rapid , violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure (caused by shot1 circuits, g rounding , arcing , etc.) should not automatically be considered an explosion.
Such events may require a post-event inspection to determine if the attributes of an expiosion are present. FIRE: Combustion characterized by heat and light. Sources of smoke such as s lipping drive belts or overheated electrical equipment do not constitute FlRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. SAFETY SYSTEM: A system required for safe plant operation , cooling down the plant and/or placing it in the cold shutdown condition , including the ECCS. These are typically systems classified as safety-related.
VTSTBLE DAMAGE: Damage to a component or structure that is readily observable without measurements , testing , or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM , or a structure containing SAFETY SYSTEM components , needed for the current operating mode. This 58 condition significantly reduces the margin to a loss or potential loss of a fission product barrier , and therefore represents an actual or potenti a l substantial degradation of the level of safety of the plant. The first threshold for EAL +CA6.1.,.lr.+ addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.
The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. The second threshold for EAL +CA6. l.,.bd addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone , or to a structure containing SAFETY SYSTEM components.
Operators will make this determination based on the totality of available event and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC CS 1 or RS 1. MNGP Basis Reference(s):
1. Ops Man A.6 (ACTS OF NATURE) 2. Ops Man B.05.14 (SEISMIC MONITORING SYSTEM) 3. Ops Man B.05.16-01 (METEOROLOGICAL MONITORING-FUNCTION & GENERAL DESCRIPTION OF SYSTEM) 4. Ops Man B.06.04 (CIRCULATING WAT E R SYSTEM) 5. C.4-B.05.14.A (EARTHQUAKE)
6. C.6-006-C-08 (EARTHQUAKE)
7. C.6-006-C-13 (OPERATIONAL BASIS EARTHQUAKE)
8. USAR Section 10.3, Plant Auxiliary Systems -Plant Service System s 9. USAR Section 12.2 , Plant Principal Structures and Foundation s 10. USAR Appendix G, Chapter 3 , Probabl e Maximum Flood Determination
11. USAR Table I.5-1 , Location of High Energy Systems and Safe Shutdown Equipment by Volume 12. USAR Appendix J.4, Fire Protection Pro g ram -Safe Shutdown Analysis 13. ND-95208, Monticello Property Map 14. ND-95209, Monticello Main Plant Structures
15. 4 AWI-01.03.01 (QUAUTY ASSURANCE PROGRAM BOUNDARY) 59 CU1 ECL: Notification of Unusual Event Initiating Condition:
UNPLANNED loss of RPV inventory for 15 minutes or longer. Operating Mode Applicability:
Cold Shutdown , Refueling Emergency Action Levels: (CU 1.1 or CU 1.2) Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded , or wi II I ikely be exceeded.
CUl.l UNPLANNED loss of reactor coolant results in RPV level less than a procedurally required lower limit for 15 minutes or longer. CUI.2 a. RPV level cannot be monitored.
Basis: AND b. UNPLANNED increase rise in drywell floor or equipment drain sumpspeeifie sump and,lor tank) levels. UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV 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 decrease RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. EAL #CUl.1 recognizes that the minimum required RPV level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented.
This EAL is met if the minimum level , specified for the current plant conditions , cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.
The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level. EAL #CU 1.2 addresses a condition where all means to determine RPV level have been lost. In this condition , operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. 60 Contin ued loss of RCS inventory may result in escalation to the Alert emergency classification level v ia either IC CAI or CA3. MNGP Basis Refere n ce(s): 1. Ops Man B.01.01-06 (REACTOR AND VESSEL ASSEMBLY -FIGURES) 2. Ops Man C.3 (SHUTDOWN PROCEDURE)
3. 9001 (REACTOR WELL AND DRYER-S E PARATOR STORAGE POOL FILLING PROCEDURE)
4. 9006 (REACTOR WELL AND DRYER-SEPARATOR STORAGE POOL DRATNTNG PROCE D URE) 5. 904 0 (TEMPORARY VESSEL LEVEL INSTRUMENTATION INSTALLATION AND RESTO RA TJON) 6. C.5-l l OO(RPVCONTROL)
7. Ops Man C.4-B.04.01.F (LEAK INSIDE PRIMARY CONTAINMENT)
8. Ops Ma n C.6-003-A-38 (REACTOR LOW LEVEL) 9. Ops Man C.6-004-B-13 (D R YWELL EQU IP D R AIN LEAK RATE H I) 1 0. Ops Man C.6-004-B-17 (DRYWELL FLOOR DRA I N SUMP HI LEVEL) 1 1. Ops Man C.6-004-8-18 (DRYWELL EQU I P DRAIN LEAK RA TE CHANGE H T) 12. Ops Man C.6-065-A-06 (REFUELING HIGH LEAKAGE) 1 3. Tec hn ica l Spec i ficat i on 3.9 (REFUELING OPERAT I ONS) 1 4. MNG P Ca l cu l atio n CA-95-0 7 4, Low Low Reactor Wate r Level Group 1 and 3 Co n ta in ment Iso l at i o n Setpo in t Ca l cu l at i o n 61 CU2 ECL: Notification of Unusual Eve nt Initiating Condition:
Loss of all but one AC power source to emergenC)' essential buses for 15 minutes or longer. Operating Mode Applicability:
Co l d Shutdown, Refueling , Defueled Emerge n cy Action Levels: Note: The Emergency Director should declare the Unusual Eve nt promptly upon determinin g that 15 minutes has been exceeded , or wi ll likely be exceeded.
CU2.1 a. AC power capabi l ity to essential buses 15 and 16 (site specific emergency buses) is reduced to a single power source (Table S 1) for 15 minutes or longer. Basis: AND b. Any additional sing l e power source failure will result in loss of a ll AC power to SAFETY SYSTEMS. Table Sl 1 R Reserve Transformer 1 AR Reserve Transformer 2R Auxiliary Transformer
11 Emergency Diesel Generator
12 Emergency Diesel Generator SAFETY SYSTEM: A system requ i red for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
This IC describes a sign i ficant degradation of offsite and onsite AC power sources (Table SI) such that any additiona l single failure would result in a loss of al 1 AC power to SAFETY SYSTEMS. In this condition , the sole AC power source may be powering one , or more than one, train of safety-re l ated equipment.
When in the cold sh u tdown , refueling, or defueled mode , this condition i s not c l assified as an Alert because of the increased time available to restore another power source to se rvice. Additiona l time is available due to the reduced L:Ore decay heat l oad , and the lower temperatures and pressures in various plant systems. Thus, when in these modes , this condition is considered to be a potential degradation of the level of safety of the plant. An "A C power source" is a source recognized in AOPs and EOPs , and capable of supplying required power to an emergency essential bus. Some examples of this condition are presented below.
A loss of all offsite power with a concurrent fai l ure of a l l but one emergency power source (e.g., an onsite diesel generator).
62
A loss of all offsite power and loss of all emergency power sm:1Fce s (e.g., onsite diesel generators) with a single traiH of ernergeHcy busesesseHtial buses being back fed from the unit main generator.
A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses essential buses being back-fed from an 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 in accordance with IC CA2. MNGP Basis Reference(s):
1. USAR Section 8.2, P l ant Electrical Systems -Transmiss i on System 2. USAR Section 8.3 , Plant E l ectrical Systems -Auxiliary Power Systems 3. USAR Section 8.4 , Plant Electrical Systems -Plant Standby Diesel Generator System 4. USAR Figure 8.4-1 , Diesel Generation System One Line Diagram 5. NF-36175 , Single L in e Diagram -Station Connection
6. Ops Man B.09.06-01 (4.16 KV STATION AUXILIARY -FUNCTION AND GENERAL DESCRIPTION OF SYSTEM) 7. Ops Man C.4-B.09.02.A (STATTON BLACKOUT)
8. Ops Man C.4-B.09.02.B (LOSS OF NORMAL OFF-SITE POWER) 9. Ops Man C.4-B.0 9.0 6.C (LOSS OF BUS 15 OR BUS 1 6) 10. Tech Spec 3.8.2 (AC SOURCES -SHUTDOWN) 63 CU3 ECL: Notification of Unusual Event Initiating Condition:
UNPLANNED increase rise in RCS temperature.
Operating Mode Applicability:
Cold Shutdown , Refueling Emergency Action Levels: (CU3.1 or CU3.2) Note: The Emergency Dir ector s hould declare the Unus ual Event promptly up on determining that 15 minutes ha s been exceeded , or will likel y be exceeded.
CU3.l UNPLANNED increase rise in RCS temperature to greater than (site specific Technical Specification cold shutdovm temperature limit)212 °F. CU3.2 Loss of ALL RCS temperature an d RPV le ve l indication for 15 minute s or lon ger. Basis: UNPLANNED:
A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold s hutdown temperature limit , or the inability to determine RCS temperature and level , represents a potential de gra dation of the le ve l of safety of the plant. If the RCS is not intact and GO:NTAlNMfil>JT GLOSURE SECONDARY CONTAINMENT is not established durin g this event , the Emergency Director should also refer to IC CA3. A momentary UNPLANNED excursion above the Technical Specification cold sh utdown temperature limit whe n the heat removal function i s available does not warrant a classification.
EAL #CU3. l in vo lves a lo ss of decay heat removal capability , or an addition of heat to the RCS in excess of that which can currently be removed, such that react or coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications.
During this condition, there is no immediate threat of fuel d a ma ge because the core deca y h eat load has been reduced since the cessation of power operation.
During an outage, the level in the reactor vesse l 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 rem ova l at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown. EAL #CU3.2 reflects a condition where there has been a significant Joss of in strume ntation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters neces sa ry to assure core decay heat removal. During this condition , there is no immediate threat of fuel dam age because the core deca y heat lo a d has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary l osses of indication.
64 Escalation to Alert would be via IC CAI based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
MNGP Basis Reference(s):
1. C.5.l-1100 (RPV CONTROL) 2. Ops Man B.01.01-03 (REACTOR AND VESSEL ASSEMBLY -INSTRUMENTATION AND CONTROLS)
3. Ops Man C.3 (SHUTDOWN PROCEDURE)
4. 9040 (TEMPORARY VESSEL LEVEL INSTRUMENTATION INSTALLATION AND RESTORATION)
5. Tech Spec Table 1.1-1 (MODES) 65 CU4 ECL: Notification of Unusual Event Initiating Condition:
Loss of Vital DC power for 15 minutes or l onger. Operating Mode Applicability:
Cold Shutdown, Refueling Emergency Action Levels: Note: The Emerge nc y Direct or shou ld declare the Unusual Eve nt promptly up o n determinin g that 15 min ut es has been exceeded , or will likel y be exceeded. CU4.l Indicated vo lt age i s less than 110 VD C(site speeifie bus voltage value) on requ ir ed 125 VDC Vital DC bu ses fo r 15 minute s or longer. Basis: This IC addresses a l oss of Vital DC power which compromises the ability to monitor and contro l operable SAFETY SYSTEMS when the plant is in the cold shutdown or refue lin g mode. In these modes , the core deca y heat load has been significantly reduced, a nd coolant system temperatures and pressures are lower; these conditions increa se the time available to restore a vital DC bus to serv ice. Thus , this condition i s con s idered to be a potent i al de gra dation of the l evel of safety of the plant. As used in this EAL, " requir ed" means the Vital DC buse s nece ssary to support operat i on of the in-service , or o perable , train or trains of SAFETY SYSTEM equipment.
For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-se r v ice (operable), then a l oss of Vital DC power affecting Train B wou ld require the declaration of an Unusua l Event. A l oss of Vital DC power to Train A wou ld not warrant a n emergency classification.
The indicated voltage used in this threshold is based on battery sizing calculations.
The threshold is an average for both Division I and TT batteries for battery voltages at 15 minutes prior to reaching the minimum required terminal voltage. The Division I and 11-250 VDC battery systems need not be considered in this EAL because they supply power to large motor loads in the RCTC and HPCI systems and various non-critical loads. RCJC is an alternative source of make-up water for the reactor during normal plant shutdowns and transient events which lead to a l oss of feedwater flow. HPCT is part of the Emergency Core Cooling System (ECCS) network. However , the Auto Depressurization System (ADS) is redundant in function to the HPCI system and does not require 250 VDC for operations.
Therefore, these systems need not be included in this EAL since loss of the 250 VDC battery systems would not cause core uncovering or loss of containment integrity.
Fifteen minutes was selected as a threshold to exc lude transient or momentary power lo sses. Depending upon the event, escalat i on of the emergency c la ssification level would be via IC CAl or CA3, or an IC in Recogniti o n Category R. 66 MNGP Basis Reference(s):
1. USAR Section 8.5.1, Plant Electrical Systems -DC Power Supply Systems , Essential 250 Vdc System 2. USAR Section 8.5.2, Plant Electrical Systems -DC Power Supply Systems , 125 Vdc System 3. NE-36640-2 , 125VDC Distribution Electrical Scheme 4. MNGP Calculation CA-02-179 , 125 Volt Div. I Calculation
5. MNGP Calculation CA-02-192 , 125 Volt Div. IT Calculation
6. Technical Specification 3.8.5 (DC SOURCES -SHUTDOWN) 67 ECL: Notification of Unusual Event Initiating Condition:
Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability:
Cold Shutdown , Refueling , Defueled Emergency Action Levels: (CU5.l or CU5.2 or CU5.3) CU5.1 Loss of ALL of the following onsite communication methods: (site speeifie list of eommBAieatioAs methods)
Commercial Telephones
Plant Telephones
Portable radios
Plant PA System CU5 CU5.2 Loss of ALL of the following Offsite Response Organization (ORO) communications method s: * (site speeific list of eommlmieatioAs methods) Commercial Telephones
Direct Dedicated Telephones
Radio/Receiver Transmitter CU5.3 Loss of ALL of the following NRC communications methods: Basis: * (site specific list of eommuHieatioAs methods) Federal Telecommunications System (FTS)
Commercial Telephones This IC addresses a significant loss of on-site or offsite communications capabilities.
While not a direct challenge to plant or personnel safety , this event warrants prompt notifications to OROs and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant , privately owned equipment , relaying of on-site information via individuals or multiple radio transmission point s, individuals being sent to offsite locations , etc.). EAL #CU5.l addresses a total loss of the communications method s used in s upport of routine plant operations.
EAL #CU5.2 addresses a total loss of the communications methods used to notify all ORO s of an emergency declaration.
The OROs referred to here are (see De11*eloper State of Minnesota , Wright County , and Sherburne County. EAL #CU5.3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
68 MNGP Basis Reference(s):
1. USAR Section 10.3.8 , Plant Auxiliary Systems -Plant Service Systems, Plant Communications System 2. MNGP Emergency Plan Section 7.2-Communication Systems 3. MNGP Emergency Plan Figure 13.7 -Direct Dedicated Telephones (Hot Lines) 4. A.2-504 Emergency Communicator Duties in the TSC and OSC 69 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS Table E 1: Reeognition Category "E" Initiating Condition Matrix UNUSUAL EVENT E-II Ul Damage to a loaded cask CONFINEMENT BOUNDARY.
Op. Modes: All 70 ISFSI MALFUNCTION E-M U1 ECL: Notification of Unusual Event Initiating Condition:
Damage to a loaded cask CONFINEME T BOUNDARY.
Operating Mode Applicability:
All Emergency Action Levels: EUl .1 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated b y an on-contact radiation reading greater than (2 times the site speeifie eask speeifie teehnieal speeifieation allm.vable radiation le't*e l) on the surface of the spent fuel eask any of the values listed in Table El: Table El Location of Dose Rate Total Dose Rate (Neutron+
Gamma mR/hr) HSM or HSM-H Front Surface 1400 HSM or HSM-H Door Centerline 2 00 End shield wall exterior 4 0 Basis: CONFINEMENT BOUNDARY:
The barrier(s) between areas containing spent fuel and the environment once the spent fuel is processed for dry storage. This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual re l ease path to the environment , degradation of one or more fuel assemblies due to environmental factors , and configuration changes which could cause challenges in removing the cask or fuel from storage. The existence of "damage" is determined by radiological survey. The technical specification multiple of"2 times", which is also used in Recognition Category RIC RUl , is used here to distinguish between non-emergency and emergency conditions.
The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask , the fact that the " on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask. Security-related events for ISFSis are covered under ICs HUI and HAI. MNGP Basis Reference(s):
1. 9508 (DSC TRANSFER FROM TRANSFER CASK TO HSM) 71 ISFSI MALFUNCTION
2. ISFSI Tech Spec 1.2.7f, HSM or HSM-H Dose Rates with a loaded Type 1 61BTH DSC Only -Amendment 10 72 9 FISSION PRODUCT BARRIER ICS/EALS Table 9 F I: Recognition Category " F" Initiating Co ndition Matrix GENERAL EMERGENCY Loss of any two barriers and Loss or Potential Loss of the third barrier. FGl Op. Modes: Power Operation, Het Str:mrih)*, Startup , Hot Shutdown SITE AREA EMERGENCY Loss or Potential Loss of any two barriers.
FSl Op. Modes: Power Operation, Het Str:mrihy, Startup , Hot Shutdown ALERT Any Loss or any Potentia l Loss of either the Fuel Clad or RCS barrier. FAl Op. Mod es: Pow e r Operation , Het Str:mrihy , Startup , Hot Shutdown See Table 9 F 2 for BWR E1\Ls LOSS POTENTIAL LOSS FUE L CLAD L OSS POTENT I AL LOSS FUEL CLAD LOSS POT E NT I A L LOSS FUEL CLAD 73 LOS S POTEN TI AL LOSS RCS POT E NTIA L LOSS Y ES IG.l -Loss or ANY Two Loss o r Poteniia.I Loss of Third Barrier LO SS POTENTIAL LOSS CON TA I NM EN T ffil -Loss or Potential Loss o f ANY T wo B.miers Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FGlGENERALEMERGENCY FSl SITE AREA EMERGENCY FAlALERT Los s of any two b a rri e rs a nd Lo ss o r Loss o r P o t e ntial L oss of a ny tw o barri e r s. A n y Loss or a n y Pot e ntial Loss of e ith e r th e Pot e ntial L oss o f th e third b a rrier. F u e l C l a d or RCS b a rri e r. Fuel Clad Barrier RCS Barrier Containment Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 1. RCS Activity 1. Primary Containment Pre s sure 1. Primary Containment Conditions A. Egite specit'ic Not Ap pli ca bl e A. Pr im ary Not App li ca bl e A. UNPLANNE D A. Prim a r y in d ications that c o nt a inm e nt ra pid d ro p in c o nt a inm e nt reactor c C o ol a nt pr ess ure g r e at e r primar y pr ess ure g r ea t e r ac ti v it y is grea ter th a n (site specific co n ta inm e nt th a n (site-th a n 3 00 µC i/gm -vaktej l .84 psig pr ess ur e fo ll ow in g specific d ose e qui va l e n t I-du e t o R CS prim a r y 131 1. l e ak age. co nt a inm e nt OR pr ess u re ri se --B. Greater than OR or equa l to 6% B. P rim ary hydrogen and co nt a inm e nt greater than or pr ess ur e r es p o n se equal to 5% n o t co n s i s t e nt w ith oxygen m L O CA co nditi o n s. Drywell or Torus (site-specific e*plosi*r<e mi*1ttrej e*ists inside primaFJ' containment OR C. H C+L exceede d. 2. RPV Water Level 2. RPV Water Level 2. RPV Water Level 7 4
-F u e l C l ad B a r rier RCS Barrier Con t ainme n t B ar r ier L O SS POTEN TI AL LOSS LOSS POTENTIAL LOSS LOSS POTENT I AL LOSS A. PFimaFJ' A. RPV water l evel A. RPV water level Not Applicable Not Applicable A. PFimaFJ' containment cannot be restored cannot be restored containment flooding SAMG and maintained and maintained flooding SAMG entry i s required.
above fsite-a bove fsite-entry is specific RPV SJ3ecific RPV required. >+vateF le¥el 'NateF le,.*el coffes13onding to coFFesponaing to the top of active the top of active fue.11-126in.or fue.11-126in.or cannot be cannot be detennined.
determined. 3. Not Applicable
3. RCS Leak Rate 3. Primary Containment Iso l ation Fa il ure Not Applicable Not Appl i cab l e A. UNISOLABLE A. UNISOLABLE A. UNISOLABLE Not Applicable break in ANY of primary system direct downstream the following:
leakage that pathway to the Esite specific results in environment exists systems '+vith exceeding after primary 13otential feF high Control Room containment eneFgy line indication of isolation signal bFeaks) MSL; EITHER of the OR HPCI;RWCU; following:
B. Intentional RCIC as indicated
1. Max Normal pnmary by high Operating containment flow/temperature Temperature venting per EO Ps isolation setpoints OR OR OR 2. Max Normal C. UNISOLABLE B. E mergency RPV Operating Area primary syste m Depressurization. Radiation leakage t h at Leve l. re s ults in exceed i ng Control Room indication of EITHER of 75 Fuel Clad Barrier RCS Barrier Containment Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS the following:
1. Max Safe Op erat in g Tern perature.
OR 2. Max Safe Operating Area Radiation Leve l. 4. Primary Containment Radiation
4. Primary Containment Radi a tion 4. Primary Containment Radiation A. Containment High Not Ap plic ab l e A. Containment High Not Applicable Not App li cab l e A. Containment High Range Rad Range Rad Range Rad (Drywell (Drywell (Drywell Radiation)
Radiation) monitor Radiation) monitor reading reading greater monitor reading greater than than greater than 1.5 E +03 R/hr 6.2 E+O l R/hr 3.3 E+04 R/hr PFimaFj' PFimaFj' containment PFimaFj' containment containment rnaiation monitoF rnaiation monitor rnaiation monitoF reaaing greateF reaaing greater reaaing greater than (site specific than (site specific than (site specific Yalue). Yalue). *ral1:1e). s. QtheF ledieatiees
s. QtheF IBdieatiens
s. QtheF ledieatiees A. (site specific ApplicaBleA.
ApplicableA.
Not ApplicableA.
ApplicableA.
Not ApplicableA.
as (site specific (site specific (site specific as (site specific (site specific Applicable as applicable) as applicable) applicable) as applicable) as applicable) 6 5. Emergency Director Judgment 6 5. Emergency Director Judgment 6 5. Emergency Director Judgment A. ANY co nditi o n in A. ANY cond iti o n in A. ANY co nd it i o n in A. ANY cond iti on in A. ANY cond iti on in A. ANY condition in the op ini o n of the the opin i on of the the opin i on of the the op ini on of the the op in ion of the t h e op ini on of the E m erge nc y E m ergency Emerge n cy Eme r gency Emergency E m erge nc y 76 F u e l C l ad B arrier RCS Bar r ier a Containment Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS Director that Director that Director that Director that Director that Director that indicates Loss of indicates Potential indicates Loss of indicates Potential indicates Loss of indicates Potent i a l the Fuel Clad Loss of the Fuel the RCS Barrier. Loss of the RCS the Containment Loss of the Barri e r. Clad Barrier. Barrier. Barrier. Containment Barrier. 77 Basis Information For BWR EAL Fission Product Barrier BWR-FUEL CLAD BARRIER THRESHOLDS:
The Fuel C l ad barrier consists of the zircalloy or stainless steel fuel bund l e tubes that contain the fuel pellets. 1. RCS Activity Loss I.A This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred , it represents a loss of the Fuel Clad Barrier. There is no Potential Loss threshold associated with RCS Activity.
2. RPV Water Level Loss 2.A The Loss threshold represents the-any EOP requirement for primary containment flooding.entry into Severe Accident Management Guidelines.
This is identified in the BWR GG EPGs/SAGs when the phrase , " Primary Containment f l ooding Js Required ," appears. Since a site specific RPV water le*1el is not specified here , the boss threshold phrase , " Primary containment flooding required ," also accommodates the EOP need to flood the primary containment when RPV water level cannot be determined and core damage due to inadequate core cooling is believed to be occurring adequate core cooling cannot be assured. Potential Loss 2.A This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling. The RPV water level threshold is the same as RCS barrier Loss threshold 2.A. Thus , this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency.
This threshold is considered to be exceeded when, as specified in the site-specific EOPs , RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually , automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization.
EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events , elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads 78 of available injection sources. Therefore , this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized , or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss ofRPV inventory.
The term " cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel , but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
In high-power A TWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL).
Although such action is a challenge to core cooling and the Fuel Clad barrier , the immediate need to reduce reactor power is the higher priority.
For such events, ICs SAS or SSS will dictate the need for emergency classification.
Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier , a potential los s of the fuel clad barrier is specified.
3. Not Applicable (included for numbering consistency between barrier tables) 4. Primary Containment Radiation Loss 4.A The radiation monitor reading corresp o nds to an instantaneous release of all reactor coolant mass into the primary containment , assuming that reactor coolant activity equals 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to S% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold 4.A since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency.
There i s no Potential Loss threshold associated with Primary Containment Radiation.
5. Other Indieations l>fot Applieable Loss and/or Potential Loss 5.A 79 J This subcategory addresses other site specific thresholds that may be included to indicate loss or potential loss of the Fuel Clad barrier based on plant specific design characteristics not considered in the generic guidance.
6.5. Emergency Director Judgment Loss {;5.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad Barrier is lost. Potential Loss {;5.A This threshold addresses any other factors that may be used by the Emergency Director in determining w h ether the F u e l Clad Barrier is potentially l ost. The Emergency Director should also consider whether or not to declare the barrier potentia ll y lost in the event that barrier status cannot be monitored.
80 BWR RCS BARRIER THRESHOLDS:
The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant s y stem piping up to and including the isolation valves. 1. Primary Containment Pressure Loss l.A The (site specific value) 1.84 psig primary containment pressure is the drywell high pressure setpoint which indicates a LOCA by automatically initiating the E CCS or equivalent makeup system. There is no Potential Loss threshold associated with Primary Containment Pressure.
2. RPV Water Level Loss 2.A This water level corresponds to the top of active fuel and is used in the EOPs to indicate challenge to core cooling. The RPV water level threshold is the same as Fuel Clad barrier Potential Loss threshold 2.A. Thus , this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency.
This threshold is considered to be exceeded when , as specified in the site-specific EOPs , RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually , automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events , elevated RPV pressure may prevent restoration of RPV water level unti I pressure drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized , or required emergency RPV depressurization has been attempted , giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.
The term, "cannot be restored and maintained above ," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
81 In high-power A TWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL).
Although such action is a challenge to core cooling and the Fuel Clad barrier , the immediate need to reduce reactor power is the higher priority.
For such events , ICs SAS or SSS will dictate the need for emergency classification.
There is no RCS Potential Loss thre s hold associated with RPV Water Level. 3. RCS Leak Rate Loss Threshold 3.A Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated.
If it is determined that the ruptured line cannot be promptl y isolated from the Control Room , the RCS barrier Loss threshold is met. Loss Threshold 3.B Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emerge ncy RPV Depressurization is performed , the plant operators are directed to open safety relief valves (SR Vs) and keep them open. Even though the RCS is being vented into the suppression pool , a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.
Potential Loss Threshold 3.A Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC , HPCI , etc., which indicate a direct path from the RCS to areas outside primary containment.
A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated suppott and control systems functioning properly.
The indicators reaching the threshold barrier s and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification.
A primar y system is defined to be the pipes , valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. An UNISOLABLE leak which is indicated by Max Normal Operating values escalates to a Site Area Emergency when combined with Containment Barrier Loss threshold 3.A (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.
82
4. Primary Containment Radiation Loss 4.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 4.A since it indicates a loss of the RCS Barrier only. There is no Potential Loss threshold associated with Primary Containment Radiation.
S. Other ludieetiens Not Applieable Loss and/or Potential boss 5.A This subeategory addresses other site speeifie thresholds that may be ineluded to indieate loss or potential loss of the RCS barrier based on plant speeifie design eharaeteristies not eonsidered in the generie guidanee.
6.5. Emergency Director Judgment Loss 65.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost. Potential Loss 6 5.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
83 BWR CONTAINMENT BARRIER THRESHOLDS:
The Primary Containment Barrier includes the drywell , the wetwell , their respective interconnecting paths , and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
1. Primary Containment Conditions Loss 1.A and 1.B Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of primary containment integrity.
Primary containment pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus , primary containment pressure not under these conditions indicates a loss of primary containment integrity.
These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned.
The unexpected (UNPLANN E D) response is important because it is the indicator for a containment bypass condition.
Potential Loss 1.A The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure.
A pressure of this magnitude is greater than those expected to result from any design basis accident and , thus , represent a Potential Loss of the Containment barrier. Potential Loss 1.B If hydrogen concentration reaches or e xceeds the lower flammability limit , as defined in plant EOPs , in an oxygen rich environment , a potentially explosive mixture exists. The existence of an explosive mixture (2: 6% f-1 2 and 2: 5% 02) means , at a minimum , that the containment hydrogen concentration is sufficient to support a hydrogen burn. If the combustible mixture ignites inside the primary containment , loss of the Containment barrier could occur. Potential Loss 1.C The Heat Capacity Temperature Limit (HC T L) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:
Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized , OR 84
Suppression chamber pressure above Primary Containment Pressure Limit A , while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent. The HC+L is a function of RPV pressure , suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore , the inability to maintain plant parameters below the limit constitutes a potential loss of containment.
2. RPV Water Level There is no Loss threshold associated with RPV Water Level. Potential Loss 2.A The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold 2.A. The Potential Loss requirement for Primary Containment Flooding entry into the Severe Accident Management Guidelines indicates adequate core cooling cannot be restored and maintained assured and that core damage is possible.
BWR EPGs/SAGs specify the conditions that require primary containment flooding.
When primary containment flooding is required , when the EPGs are exited and SA M Gs are entered. Entry into S AM Gs is a logical escalation in response to the inability to restore and maintain assure adequate core cooling. PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which , if not corrected , could lead to RPV failure and increased potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns , this threshold results in the declaration of a General Emergency.
3. Primary Containment Isolation Failure These thresholds address incomplete containment isolation that allows an UNISOLABLE direct release to the environment.
A re l ease path is 'direct' if it allows for the migration of radioactive material from the containment to the environment in a generally uninterrupted manner (e.g., littl e or no holdup time); therefore , within the context of a Containment Barrier Loss or Potential Loss thre s hold , a release path through the wetwell is a direct release path. Loss 3.A The use of the modifier " direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways , such as instrument lines , not protected by the Primary Containment Isolation System (PCIS). The existence of a filter is not considered in the threshold assessment.
Filters do not remove fission product noble gases. In addition , a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream. 85 Following the leakage of RCS mass into primary containment and a rise in primary containment pressure , there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components.
Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R ICs. Loss 3.B EOPs may direct primar y containment isolation valve logic (s) to be intentionally bypassed , even if offsite radioactivity release rate limits will be exceeded.
Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed.
Intentional venting of primary containment for primary containment pre ss u re or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment.
In this situation conditions and trends are such that the Control Room staff has made a decision to perform an intentional controlled venting of the containment.
This intentional venting action results in a bypass of the primary containment, whether it is anticipatory or otherwise.
Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the drywell hi g h pressure scram setpoint) does not meet the threshold condition.
Loss 3.C The Max Safe Operating Temperature and the Max Safe Operatin g Radiation Level are each the hi g hest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded.
EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization i s required.
The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes , valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. In combination with RCS potential loss 3.A this threshold would result in a Site Area Emergency.
There is no Potential Loss threshold associated with Primary Containment Isolation Failure. 4. Primary Containment Radiation There is no Loss threshold associated with Primary Containment Radiation. Potential Loss 4.A 86 The radiation monitor reading corresponds to an instantaneous re l ease of all reactor coolant mass into the primary containment , assuming that 20% of the fuel cladding ha s failed. Th i s le ve l of fuel clad fa ilur e i s well above that used to determine the ana l ogous Fuel Clad Barrier Loss and RCS Barrier Loss thresho ld s. NUREG-1228 , Source Estimations During Incid e nt Response to Severe Nu clear Power Plant Accidents, indicates the fuel clad fai lu re must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offs it e protective actions. For this condit ion to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment wh i ch would then esca l ate the emerge nc y classification l evel to a General Emergency.
5. Other Indieations Not Applieable Loss and/or Potential Loss 5.A This subeategory addresses other site speeifie thresholds that may be ineluded to indieate loss or potential loss of the Containment barrier based on plant speeifie design eharaeteristies not eonsidered in the generie guidanee.
9.5. Emergency Director Judgment Loss 6 5.A This threshold addresses any other factors that are to be u se d by the Emergency Director in determining whether the Containment barrier is lo st. Potential Loss 65.A T hi s threshold addresses any other factors that may be used by the Emergency Director in determining whether the Co nt a inm ent Barrier is potentially lost. The Emerge n cy Director shou ld a l so cons ider whether or not to declare the barrier potentially l ost in the event that barrier status ca nn ot be monitored.
MNGP Basis Reference(s):
1. Ops Man B.02.04 (MAIN STEAM) 2. Ops Man C.4-B.04.01.A (PRIMARY CONTAINM E NT ISOLATION -GROUP 1) 3. Ops Man C.6-005-A-25 (MATN STEAM LTNE HI FLOW CH A) 4. Ops Man C.6-005-A-26 (MAIN STEAM LINE Hl FLOW CH B) 5. Ops Man C.6-003-8-56 (HIGH AREA TEMP STEAM LEAK) 6. Ops Man C.5.1-1100 (RPV CONTROL) 7. Ops Man C.5.1-1200 (PRIMARY CONTAINMENT CONTROL) 8. Ops Man C.5.1-1300 (SECONDARY CONTAINMENT CONTROL) 87
9. Ops Man C.5.1-2006 (RPV FLOODING)
10. Ops Man C.5.1-2007 (FAILURE TO SCRAM) 11. C.5-3505 (VENTING PRIMARY CONTAINMENT)
12. MNGP Calculation CA-04-194, Containment High Range Radiation Monitor (CHRRM) Response to Drywell Activity 13. USAR Section 5.2 , Containment System -Primary Containment System 14. USAR Table Section 7.5 , Plant Instrumentation and Control Systems -Plant Radiation Monitoring Systems 15. Tech Spec Table 3.3.3.1-1 (Post Accident Monitoring Instrumentation)
16. Tech Spec 3.6.1.1 (Primary Containment)
17. Tech Spec 3.6.4.1 (Secondary Containment)
18. NX-7831-197-1 , Reactor Vessel & Internals 88 10 H AZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFE T Y ICS/EALS Table H 1: Reeognition Category "H" Initiating Condition Matrix GENERAL SITE AREA ALERT UNUSUAL EVENT EMERGENCY EMERGENCY H Gl HOSTILE HSl H OSTIL E HAl HOSTILE HUl Confirmed ACTION resultin g in ACTION within the ACTION within the SECURITY loss of ph ys ical control Plant PROTECTED OWNER CONDITION or threat. of the facility. AREA. CONTROLLED AREA Op. Modes: A ll Op. Modes: A ll Op. Modes: A ll or airborne attack threat within 30 minute s. Op. Modes: All HU2 Seismic event greater than OBE leve l s. Op. Modes: A ll HU3 Hazardous event. Op. Modes: A ll HU4 FIRE potent i ally de grad ing the level of safety of the plant. Op. Modes: A ll HAS Gaseous release i mpedin g access to equipment nece ssary for normal p l ant operations , cooldown or shutdown.
Op. Modes: All HS6 Inability to HA6 Contro l Room contra I a key safe ty evacuation resultin g in function from outside transfer of plant contra 1 the Control Room. to alternate locati ons. Op. Modes: A ll Op. Modes: All HG7 Other conditions HS7 Other conditions HA7 Other conditions HU7 Other conditions ex i st which in t h e exist which i n the ex i st which in the exist which in the judgment of the judgment of the jud g ment of the jud gment of the Emergency Director Emergency D i rector E mergency Director Emergency Director warrant declaration of a warrant dec l arat i on of a warrant declaration of warrant declaration of a General Emergency.
Site Area Emergency.
an Alert. (NO)UE. Op. Modes: All Op. Modes: A ll Op. Modes: All Op. Modes: A ll 89 HG1 ECL: General Emergency Initiating Condition:
HOSTILE ACTION resulting in loss of physical control of the facility. Operating Mode Applicability:
All Emergency Action Levels: HG 1.1 a. A HOSTILE ACTION is occurring or has occurred within the Plant PROTECTED AREA as reported by the Sec uri ty Shift Supe r visor (site specific security shift supervision). AND b. EITHER of the following has occurred:
1. ANY of the following safety functions cannot be controlled or maintained.
Reactivity control
RPV water level [BWR]
RCS heat removal OR 2. Dama g e to spent fuel has occurred or is IMMIN E NT. Basis: HOSTILE ACTION: A n act toward a NPP or it s personne l t h at in cludes t h e use of vio l e n t force to destroy eq u ipment, take HOSTAGES, a n d/or i ntimidate t h e l icensee to ac h ieve an end. This in cl u des attack by a ir , l and , o r water u s ing g un s, exp l os i ves, PR OJECT I LEs , vehic l es , o r other dev i ces u sed to de l iver d estructive force. Other acts that satisfy the overa ll intent may be i nc lu ded. HOSTILE ACTION s h o uld not be co n strued to i n c lu de acts of c i v il disobed i e n ce or felon i o u s acts that are not part of a concerted attack on the NPP. Non-terror i sm-based EA Ls s h ould be u sed to address such act i vities (i.e., t hi s may include violent acts between i ndividuals in t h e ow n er contro ll e d area). JMMlNENT:
T h e tra j ectory of eve n ts or co n d i t i o n s is suc h t h at an EAL w ill be met w i t hin a re l at i ve ly s h ort pe ri o d of time r ega rdl ess of mi t i ga ti o n or cor r ective actio n s. P ROTECTED AREA: The area s u rro un d ing t he p l a nt e n co m passed by t h e cha i n lin k fe n ce and certai n st ru ctures as defined in t h e Sec u rity P l a n; excludes t h e I SFSI Protected Area. In areas w h ere two fences are present, the i nn er fence i s designated as the Protected Area barr i er. This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions.
It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a s pent fuel pool cooling system (e.g., pumps , heat exchangers , controls , etc.) or , 2) loss of spent fuel pool integrity such that sufficient w ater level cannot be maintained.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a securit y-related event. Security plans and terminology are based on the guidance provided by NEI 03-12 , Template for 90 the Security Plan , Training and Qualification Plan , Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate Security-sensitive information.
This includes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.
Security-sensitive information should be contained in non-public documents such as the Security Plan. MNGP Basis Reference(s):
I. MNGP Safeguards Contingency Plan 2. C.4-L (RESPONSE TO SECURITY THREATS) 3. NEI 03-12, Template for the Security Plan , Training and Qualification Plan , Safeguards Contingency Plan (and Independent Spent Fuel Storage Installation Security Program) 4. ND-95209, Monticello Main Plant Structures
5. QFl 775 (DEFINITIONS) 91 HG7 ECL: General Emergency Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency. Operating Mode Applicability:
All Emergency Action Levels: HG7.l Other conditions exist wh ich in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the faci lit y. Releases can be reasonab l y expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. Basis: HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment , take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air , land, or water using guns, explosives, PROJECTILEs , vehicles , or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). IMMINENT:
The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification l evel description for a General Emergency.
MNGP Basis Reference(s):
1. QFl 775 (DEFINITIONS) 92 HS1 ECL: Site Area Emergency Initiating Condition:
HOSTILE ACTION within the Plant PROTECTED AREA. Operating Mode Applicability:
All Emergency Action Levels: HSl.1 A HOSTILE ACTION is occurring or has occurred within the Plant PROTECTED AREA as reported b y the Security Shift Supervisor (site specific security shift supervision). Basis: HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES , and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns , explosives , PRO.JECTJLEs , vehicles, or other devices used to deliver destructive force. Other that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). PROTECTED AREA: The area surrounding the plant encompassed by the chain link fence and certain structures as defined in the Security Plan; excludes the ISFSI Protected Area. ln areas where two fences are present , the inner fence is designated as the Protected Area barrier. This IC addresses the occurrence of a HOSTILE ACTION w ithin the Plant PROTECTED AREA. This event wi ll require rapid response and assistance due to the poss ibil ity for damage to plant equipment.
Timely and accurate communications between Security Shift S up ervision and the Control R oo m is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12 , Templat e for th e Security Plan , Training and Qualification Plan , Safeguards Cont ing e nc y Plan [and Independent Spent Fuel Storage Installation Security Program}.
As time and conditions allow , these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering).
The Site Area Emergency declaration will mobilize ORO resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successfu l in impairing multiple safety functions. This IC does not app l y to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA located outside the plant PROTECTED AREA; such an attack should be assessed using IC HA 1. It also does not apply to in cidents that are accide nt al events, acts of civil disobedience , or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examp l es include the crash of a small aircraft , shots from hunters , physical disputes between employees , etc. 93 Reporting of these types of events is adequately addressed by other EALs , or the requirements of 10 CFR § 73.71or10 CFR § 50.72. Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate Security-sensitive information.
This includes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.
Security-sensitive information should be contained in non-public documents such as the Security Plan. Escalation of the emergency classification level would be via IC HG 1. MNGP Basis Reference(s):
1. MNGP Safeguards Contingency Plan 2. C.4-L (RESPONSE TO S E CURITY THREATS) 3. NEI 03-12 , Template for the Security Plan , Training and Qualification Plan , Safeguards Contingency Plan (and Independent Spent Fuel Stora g e Installation Security Program) 4. ND-95209 , Monticello Main Plant Structures
5. QFl 775 (DEFINITIONS) 94 HS6 ECL: Site Area Emergency Initiating Condition:
Inability to control a key safet y function from outside the Control Room. Operating Mode Applicability:
Power Operation , Startup , Hot Shutdown , Cold Shutdown , Refueling Al-J. Emergency Action Levels: Note: The Emergency Director shou ld declare the Site Area Emergency promptly upon determining that 10 minutes (site specific number of minutes) has been exceeded , or will likel y be exceeded.
HS6. l a. An event has resulted in plant control being transferred from the Control Room to the alternate shutdown panel (s i te specific remote shutdown panels and l ocal control stations). Basis: AND b. Control of ANY of the following key safety functions is not reestab l ished within 10 minutes (site specific number of minutes).
Reactivit y contro l (Modes 1 and 2 only)
RPV water level [BWR] *
RCS heat removal This IC addresses an evacuation of the Contro l Room that results in transfer of plant control to a lternate locations , and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain contro l of a key safety function fo ll owing a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period oftime. The determination of whether or not " control" is estab li shed at the-alternate shutdown panel remote safe shutdovm locat i on(s) is based on Emergency Director judgment.
The Emergency Director is expected to make a reasonable , informed judgment within 10 (the site specific time for transfer) minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown loc ation(s). Ops Man C.4-C , Shutdown Outside the Control Room, provides specific instructions for evacuating the Control Room and establishing plant control at the alternate shutdown panel. It should be noted here that analysis has shown that under worst case conditions (Assumed power level of 2004 MWt, Control Room fire coincident with loss of offsite power and reactor isolation and no available RPV injection) that indicated reactor level will decrease below the Top of Active Fuel (TAF) in approximately 11 minutes. Additionally , spurious operation of an SRV as a result of the fire event may lead to indicated reactor water level already being below TAF at the time of Operator arrival to the ASDS panel at the 10 minute mark. The EOPs would normally require depressurization before level reaches TAF however, thermal hydraulic analysis was performed and , given the above mentioned bounding scenarios, conservatively assumed reactor manual depressurization to occur at 17 minutes. Escalation of the emergency classification level would be via IC FG 1 or CG 1. 95 MNGP Basis Reference(s):
l. Ops Man C.4-C (SHUTDOWN OUTSIDE THE CONTROL ROOM) 96 HS7 ECL: Site Area Emergency Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency.
Operating Mode Applicability:
All Emergency Action Levels: HS7. l Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts , (1) toward s it e personnel or equipment that could lead to the likely failure of or , (2) that prevent effect i ve access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary. Basis: HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES , and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives , PROJECTILEs , vehicles , or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a conce1ied 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). This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a Site Area Emergency.
MNGP Basis Reference(s):
I. QFl 775 (DEFINITIONS) 97 HA1 ECL: Alert Initiating Condition:
HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes. Operating Mode Applicability:
All Emergency Action Levels: (HA 1.1 or HA 1.2) HAl.l A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervisor (s ite specific security shift st1pen 1 ision). HAl.2 A validated notification from NRC of an aircraft attack threat within 30 minutes of the site. Basis: HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES , and/or intimidate the licensee to achieve an end. This includes attack by air , land , or water using guns, explosives , PROJECTILEs , vehicles , or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EA Ls should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). OWNER CONTROLLED AREA: The OCA boundaries consist of the plant property enclosed by a three strand barbed wire fence and a posted boundary on the Wright County side of the river. This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the Plant PROTECTED AREA , or the need to prepare the plant and staff for a potential aircraft impact. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12 , Templat e for the Security Plan , Training and Qualification Plan , Safeguards C ontingenc y Plan [and Ind e pendent Sp e nt Fuel Storag e In s tallation Security Program].
As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation , dispersal or sheltering).
The Alert declaration will also heighten the awareness of Offsite Response Organizations , allowing them to be better prepared should it be necessary to consider further actions. This IC does not apply to incidents that are accidental events , acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include 98 the crash of a small aircraft , s hots from hunters , physical disputes between employees , etc. Reporting of these types of events is adequately addressed by other EALs , or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. EAL #HA 1.1 is applicable for any HOSTILE ACTION occurring , or that has occu r red , in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that i s located outside the plant PROTECTED AREA. EAL #HA 1.2 addresses the threat from the impact of an aircraft on the plant , and the anticipated arrival time is within 30 minutes. The intent of this EAL is to en s ure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness.
This EAL is met when the threat-related information ha s been validated in accordance with f site specific procedure s j. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.
The status and size of the plane may be provided by NORAD through the NRC. In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected , although not certain, that notification by an appropriate Federal agency to the site would clarify thi s point. In this case , the appropriate federal agency is intended to be NORAD , FBI , FAA or NRC. The emergency declaration , including one based on other ICs/EALs , should not be unduly delayed while awaiting notification by a Federal agency. Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate Security-sensitive information.
This includes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.
Security-sensitive information should be contained in non-public documents such as the Security Plan. Escalation of the emergency classification level would be via IC HS 1. MNGP Basis Referen c e(s): 1. MNGP Safeguards Contingency P l an 2. C.4-L (RESPONSE TO SECURITY THREATS) 3. NEI 03-1 2, Te m p l ate for the Sec ur ity P l a n , Tra i ning a n d Qua l ification P l an, Safeg u ards Co n t in gency P l an (a n d Indepe n de n t Spe nt F u e l Storage In sta ll ation Secur i ty Program) 4. FP-S-FSlP-08 (CONTINGENCY PLAN IMPLEMENTING PROCEDURES)
5. QF 177 5 (DEFIN I TIONS) 99 HAS ECL: Alert Initiating Condition:
Gaseous release impeding access to equipment necessary for normal plant operations , cooldown or shutdown.
Operati n g Mode Applicability:
All Emerge n cy Action Levels: Note: If the equipment in the listed room or area was already inoperab l e or out-of-service before the event occurred , then no emergency classification is warranted.
HAS.I a. AND Release of a toxic , corrosive, asphyxiant or flammable gas into any of the following Table H l plant rooms or areas: Tab l eHl Building Rooms Applicable Mode(s) Reactor Bui l ding A ll All Turbine Bui l ding A l l All Intake Structure A l l All (site speeifie list of plant rooms or areas with entry related mode applieability identified)
b. Entry into the room or area is prohibited or impeded. Basis: This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain norma l plant operation, or required for a normal plant cooldown and shutdown.
This condition represents an actua l or potential substantial degradation of the level of safety of the plant. An Alert declaration is warranted if entry into the affected room/area is , or may be, procedurally required during the p l ant operating mode in effect at the time of the gaseo u s release. T h e emergency classification is not contingent upon whether entry is actually necessary at the time of the release. Evaluation of the IC and EAL do not require atmospheric sampling; it on l y requires the Emergency Director's j u dgment that the gas concentration in the affected room/area is sufficient to precl u de or significantly impede procedura ll y requ i red access. This judgment may be based on a variety of factors inc l uding an ex i sting job hazard ana l ys i s , report of i ll effects on personnel , advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary meas u res are necessary to faci l itate entry of personnel into the affected room/area (e.g., requiring u se of protective equipment , such as SCBAs, that is not routinely emp l oyed). An emergency declaration is not warranted if any of the fo ll owing condit i ons apply. 100
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 examp l e , the plant is in Mode 1 when the gaseous release occurs , and the procedures used for normal operation , coo ld own and s hutd own do not require entry into the affected room until Mode 4.
The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).
The action for wh ich room/area entry is requ ir ed is of an administrative or record keep in g nature (e.g., norma l rounds or routine inspections).
The access contro l measures are of a conservative or precautionary nature , and wou ld not actua ll y prevent or im pede a req uir ed action. An asp h yxiant is a gas capable of reducing the l eve l of oxygen in the body to dangerous l evels. Most commonly , asphyxiants work by merely displacing air in an enclosed environment.
This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties , unconsciousness or eve n death. This EAL does not app l y to firefig htin g activit i es that automatically or manually activate a fire suppress ion system in an area , or to int entiona l inerting of containment (BWR only). Escalation of the emergency class i fication level would be via Recognition Category R , C or F I Cs. MNGP Basis Reference(s):
1. Ops Man C. l (STARTUP PROCEDURE)
2. Ops Man C.3 (SHUTDOWN PROCEDURE)
3. USAR Table 1.5-1 , Location of High Energy Systems and Safe Shutdown Equipment by Volume 4. USAR Section 10.3, Plant Auxi li ary Systems-Plant Service Systems 5. USAR Section 12.2, Plant Structures and Sh i e l ding-Plant Principal Structures and Foundations
6. ND-95209, Mont i ce l lo Main P l ant Structures 101 HA6 ECL: Alert Initiating Condition:
Control Room evacuation resulting in transfer of plant control to alternate locations.
Operating Mode Applicability:
All Emergency Action Levels: HA6.l An event has resulted in plant control being transferred from the Control Room to the alternate shutdown panel (site speeifie remote shutdown panels and loeal eontrol stations). Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety. Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations.
The necessity to control a plant shutdown from outside the Control Ro om , in addition to responding to the event that required the evacuation of the Control Room , will present challenges to plant operators and other on-shift personnel.
Activation of the ERO and emergency response facilities will assist in responding to these challenges.
Escalation of the emergency classification level would be via IC HS6. MNGP Basis Reference(s):
1. Ops Man C.4-C (SHUTDOWN OUTSfDE THE CONTROL ROOM) 102 HA7 ECL: Alert Initiating Condition:
Other cond iti ons exist which in the judgment of the Emergency Director warrant declaration of an Alert. Operating Mode Applicability:
All Emergency Action Levels: HA7.l Basis: Other conditions exist which , in the judgment of the Emergency Director , indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limit ed to sma ll fractions of the EPA Protective Action Guideline exposure levels. HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment , take HOSTAGES , and/or intimidate the licensee to achieve an end. This includes attack by a i r , land , or water using guns , explosives, PROJECTILEs , vehicles, or other devices used to deliver destructive force. Othe r acts that satisfy the overal l intent may be included.
HOSTILE ACTION s h o ul d not be co n strued to inc l ude acts of civil disobed i ence or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EA Ls should be used to address such activ i ties (i.e., this may include violent acts between individuals in the owner controlled area). This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an Alert. MNGP Basis Reference(s):
1. QFl 77 5 (DEFrNITTONS) 103 HU1 ECL: Notification of Unusual Event Initiating Condition:
Confirmed SECURITY CONDITION or threat. Operating Mode Applicability
Al I Emergency Action Levels: (HU 1. I or HU 1.2 or HU 1.3) HUI.I A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervisor (site specific secttrity shift sttpervision). HUI.2 Notification of a credible security threat directed at the site MNGP. HUl.3 A validated notification from the NRC providing information of an aircraft threat. Basis: HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES , and/or intimidate the licensee to achieve an end. This includes attack by air , land , or water using guns , explosives , PROJECTlLEs , vehicles , or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). SECURITY CONDITION:
Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security , threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION. This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment , and thus represent a potential degradation in the l evel of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of IO CFR § 73.7I or I 0 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAI , HSI and HGl. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper c l assification of a security-re l ated event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and OROs. Security plans and te rmin ology are based on the gu id ance provided by NEI 03-12 , Template for the Security Plan , Training and Qualification Plan , Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program}.
EAL #HU 1. I references (site specific seottrity shift sttpervision)
Security Shift Supervisor because #lese-this are-is the individual 5 trained to confirm that a security event is occurring or has occurred.
Training on security event confirmation and classification is controlled due to the nature of Safeguards and IO CFR § 2.39 information.
104 EAL #HUl .2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with f site specific procedure s 1. EAL #HU l .3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) wil l communicate to the licensee if the threat invo l ves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with f site specific procedure s 1. Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate Security-sensitive information. This includes information that ma y be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security P l an. Escalat i on of the emergency classification leve l would be via IC HAI. MNGP Bas i s Reference(s)
1. MNGP Safegua r ds Contingency P l an 2. C.4-L (RESPONSE TO SECURITY THREATS) 3. NEI 03-12 , Temp l ate for the Sec u rity Plan, Tra i ning a n d Qualification P l an , Safeguards Co n t in gency P l a n (a n d I ndepe nd e n t Spe n t F u e l Storage I n sta ll ation Sec u r i ty Program) 4. FP-S-FSIP-08 (CONTINGENCY PLAN I MPLEMENT I NG PROCEDURES)
5. QFl 775 (DEFTNTTTONS) 105 HU2 ECL: Notification of Unusual Event Initiating Condition:
Seismic event greater than OBE levels. Operating Mode Applicability:
All Emergency Action Levels: HU2.l Seismic event greater than Operating Basis Earthquake (OBE) as indicated by: (site specific indication that a seismic event met or exceeded OBE limit s) Annunciator OPERATIONAL BASIS EARTHQUAKE (6-C-13) received.
Basis: This IC addres ses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE)4. An earthquake greater than an OBE but l ess than a Safe Shutdown Earthquake should have no significant impact on safety-related systems, structures and components; however , some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspection s, and fully understand any impacts, this event represents a potential degradation of the l evel of safety of the plant. The Control Room annunciator EARTHQUAKE (6-C-08) alarms either by the seismic trigger of the Accelerograph Recording System or seismic switch of the Seismic Annunciator System. The annunciator OPERATIONAL BASIS EARTHQUAKE (6-C-13) alarms when it s switch senses an acceleration 2: 0.03g. The Accelerograph Recording System records accelerations in three directions, longitudinal, transversal and vertical.
This IC is based on the USAR operating basis earthquake (OBE) of 0.04g vertical or 0.06g horizontal.
Classification for this IC is to occur upon receipt of annunciator 6-C-13 as it is immediately available to CR personnel and is readily assessed.
Seismic events of this magnitude (i.e.,> OBE but:::; DBE) have been analyzed and designed for at the MNGP. However, events of this magnitude can result in plant equipment being subjected to forces that require further engineering attention.
Event verification with external so urc es shou ld not be necessary during or fol l owing an OB E. Earthquakes of this magnitude sho uld be readily fe lt by on-site personnel and recognized as a seismic event (e.g., typical lateral accelerations are in excess of 0.08g). The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however , the verification action must not preclude a timely emergency declaration. 4 An OBB is vieratery gre1:1nEI metien fer whieh these feat1:1res efa n1:1elear pewer plant neeessary fer eentin1:1eEI eperatien
,yithe1:1t unaue risk ta the health anEI safety efthe p1:1elie will remain funetional.
" An SSe is vibratory gre unEI metien fer whieh eertain (generally, safety relates) str1:1et1:1res, systems, anEI eomponents m1:1st ee ElesigneEI to ren:iain funetional.
106 Depending upon the plant mode at the time of the event , escalation of the emergency classification level would be via IC CA6 or SA9. MNGP Basis Refe r ence(s): 1. Ops Man A.6 (ACTS OF NATURE) 2. Ops Man B.05.14-01 (SEISMIC MONITORING-SYSTEM OPERATION)
3. C.4-B.05.14.A (EARTHQUAKE)
4. C.6-006-C-08 (EARTHQUAKE)
5. C.6-0 06-C-13 (OPERATIONAL BAS I S EARTHQUAKE)
6. USAR Section 2.6.5 , Seismic Mon i toring System 7. USAR Section 12.2.1 , Plant Pr i ncipal Struct ur es and Foundations , Design Basis 107 HU3 ECL: Notification of Unusual Event Initiating Condition:
Hazardous event. Operating Mode Applicability:
All Emergency Action Levels: (HU3.l or HU3.2 or HU3.3 or HU3.4 or HU3.5 or HU3.6) Note: EAL HU3.4 does not apply to routine traffic impediments such as fog , snow , ice, or vehicle breakdowns or accidents.
HU3.l A tornado strike with in the Plant PROTECTED AREA. HU3.2 Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode. HU3.3 Movement of personnel within the Plant PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chem i cal spill or toxic gas release).
HU3.4 A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.
HU3 .5 (Site specific list of natural or technological hazard events) River level g reater than 918 ft elevation.
HU3.6 River level less than 902.4 ft elevation. Basis: PROTECTED AREA: The a rea surrounding the plant encompassed b y the chain link fence and certain structures as defined in the Security Plan; exclude s the ISFSI Protected Area. In area s where two fences are present , the inner fence is designated as the Protected Area barrier. SAFETY SYST E M: A system required for s afe plant operation , cooling down the plant and/or placing it in the cold shutdown condition , including the E CCS. The s e are t y picall y system s classified as safety-related.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. EAL #HU3.1 addresses a tornado striking (touching down) within the Protected Area. EAL #HU3.2 addresses flooding of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. C l assificat ion is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay t rip). To warrant classification , operability of the affected component must be required by Technical Specifications for the current operating mode. 108 EAL #HU3.3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the Plant PROTECTED AREA. EAL #HU3.4 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site flooding caused by a hurricane , heavy rains, up-river water releases , dam failure , etc., or an on-site train derailment block i ng the access road. This EAL is not intended apply to routine impediments such as fog, snow , ice, or vehicle breakdowns or accidents , but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992 , the flooding around the Cooper Station during the Midwest floods of 1993 , or the flooding around Ft. Calhoun Station in 2011. EAL #HU3.5 addresses (site specific description) a potential flood condition.
The 918 ft elevation is selected for this EAL because it is the first elevation at which procedure actions are required to address flooding situations.
EAL HU3.6 addresses low river flow conditions. The low river water level threshold (902.4 ft elevation) corresponds to the low river flow threshold of 240 CFS. Low river level (i.e., flow) may be a precursor to loss of the ultimate heat sink and warrants further management attention.
Escalation of the emergency classification level would be based on I Cs in Recognition Categories A , F , Sor C. MNGP Basis Reference(s):
1. Ops Man A.6 (ACTS OF NATURE) 2. USAR Section 10.3, Plant Auxi l iary Systems -Plant Service Systems 3. USAR Section 12.2 , Plant Structures and Shie l ding-Plant Principal Structures and Foundations
4. USAR Appendix G , Chapter 3 , Probable Maximum Flood Determination
5. ND-95208, Monticello Property Map 6. ND-95209, Mont i ce l lo Main P l ant Structures 109 HU4 ECL: Notification of Unusual Event Initiating Condition:
FIRE potentially degrading the level of safety of the plant. Operating Mode Applicability:
All Emergency Action Levels: (HU4.1 or HU4.2 or HU4.3 or I-IU4.4) Note: The Emergency Director should declare the Unusual Event promptly upon determining that the applicable time has been exceeded , or will likely be exceeded.
HU4.l a. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications

Report from the field (i.e., visual observation)
Receipt of multiple (more than 1) fire alarms or indications
Field verification of a single fire alarm AND b. The FIRE is located within ANY of the follo*.ving Table H2 plant rooms or areas+. (site specific list of plant rooms or areas) HU4.2 a. Receipt of a single fire alarm (i.e., no other indications of a FIRE). AND b. The FIRE is located within ANY of the following Table H2 plant rooms or areas+. (site specific list of plant rooms or areas) AND c. The existence of a FIRE is not verified within 30-minutes of alarm receipt. HU4.3 A FIRE within the Plant PROTECTED AREA or ISFSI [ferplfmts with an ISFS! eiltside theplent Proteete6'Aree]
PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication.
HU4.4 A FIRE within the Plant PROTECTED AREA or ISFSI [fei*plan!s with en ISFSI eutside the plant Proteete6'Aree]
PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
Table H2 Buildin!!
Name Room(s)/Area(s) with Safety Equipment Reactor Building All HPCI Building All Turbine Bui I ding All Control and Administration Control Room , Cable Spreading Room , and Building Battery Rooms Diesel Generator Bui l ding All Diesel Fuel Oil Transfer House All EFT Building All Intake Structure All 110 Basis: FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. ISFSI (INDEPENDENT SPENT FUEL STORAGE INSTALLATION):
A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. ISFSI PROTECTED AREA: The area surrounding the Independent Spent Fuel Storage Installation encompassed by the double chain link fence surrounding the ISFSI as defined in the Security Plan; the ISFSI Protected Area is excluded from the Plant Protected Area. PROTECTED AREA: The area surrounding the plant encompassed by the chain link fence and certain structures as defined in the Security Plan; excludes the ISFSI Protected Area. In areas where two fences are present, the inner fence is designated as the Protected Area barrier. This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. EAL#HU4.1 The intent of the 15-minute duration is to size the FIRE and to discriminate against sma ll FIRES that are readily extinguished (e.g., smo lderin g waste paper basket). In addit i on to alarms , other indications of a FIRE could be a drop in fire main pressure , automatic activation of a suppression system, etc. Upon receipt , operators will take prompt act ion s to confirm the validity of an initial fire alarm , indication , or report. For EAL assessment purposes , the emergency declaration clock starts at the time that the initial alarm , indication , or report was received , and not the time that a subsequent verification action was performed.
Similarly , the fire duration clock also starts at the time of receipt of the initi al alarm , indication or report. EAL#HU4.2 This EAL addresses receipt of a single fire alarm , and the ex i stence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt , operators w ill take prompt actions to confirm the va lidit y of a sing l e fire alarm. For EAL assessment purposes , the 30-minute clock starts at the time that the initial a l arm was received, and not the time that a subsequent verification action was performed.
A sing le fire alarm, absent other indication(s) of a FIRE , may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason , additiona l time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amo unt of time to determine if an act ual FIRE exists; howev er, after that time , and absent inform ation to the contrary, it is assumed that an act u a l FIRE is in progress.
If an actual FIRE is verified by a report from the field, then EAL #HU4.1 is immediately applica ble , and the emergency mu st be declared i f the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spur iou s 111 activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.
EAL#HU4.3 In addition to a FIRE addressed by EAL #HU4.1 or EAL #HU4.2, a FIRE within the Plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA. [Senteneeferpl<:mts with en ISF'Sf eutside the plemt Pretected Aree] EAL#HU4.4 If a FIRE within the Plant PROTECTED AREA or ISFSI fferplants with cV9 JSFSI eutside #w plcmt PreteetedAree]
PROT ECTE D AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by , or supporting post-extinguishment recovery or investigation actions. Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50 , states in part: Criterion 3 of Appendix A to this part specifies that "Structures , systems, and components important to safety shall be designed and located to minimize , consistent with other safety requirements, the probability and effect of fires and explosions." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.
Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. In addition, Appendix R to 10 CFR 50 , requires, among other considerations , the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in EAL #HU4.2, the 30-minutes to verify a si ngle alarm is well within this worst-case 1-hour time period. Depending upon the plant mode at the time of the event , escalation of the emergency classification level would be v ia IC CA6 or SA9. MNGP Basis Reference(s):
I. USAR Section I 0.3, Plant Auxiliary Systems -Plant Service Systems 112
2. USAR Section 12.2 , Plant Structures and Shielding-Plant Principal Structures and Foundations
3. USAR Appendix I, Table T.5-1 , Location of High Energy System and Safe Shutdown Equipment by Volume 4. USAR Appendix J.4, Fire Protection Program -Safe Shutdown Analysis 5. ND-95209 , Monticello Main Plant Structures
6. NF-36300-1-2 , Block Wall Schedule Reactor Building 7. NF-36300-1-3 , Block Wall Schedule Turbine Building 8. NF-3600-1-4 , Block Wall Schedule Plant Admin Building and Offgas Stack 9. 4 AWI-01.03.01 (QUALITY ASSURANCE PROGRAM BOUNDARY)
10. Ops Man B.08.05-05 (FIRE PROTECTION SYSTEM OPERATION) 113 HU7 ECL: Notification of Unusual Event Initiating Condition:
Other cond i tions exist which in the judgment of the Emergency Director warrant declaration of a fN Qf UE. Operating Mode Applicability:
All Emergency Action Levels: HU7.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist wh ich are believed by the E mergency Director to fall under the emergency classification l evel description for a N Q UE. MNGP Basis Reference(s):
1. QFl 775 (DEFINITIONS) 114 11 SYSTEM MALFUNCTION ICS/EALS Table S 1: Reeognition Category "S" Initiating Condition Matrix GENERAL EMERGENCY SGl Prolonged Joss of a ll offsite and a ll onsite AC power to emergene;' essential buses. Op. Modes: Power Oper a tion , Startup, He/ Standby , Hot Shutdown SITE AREA EMERGENCY SSl Loss of a ll offsite and all onsite AC power to emergeney essential buses for 15 minutes or longer. Op. Modes: Power Operation , Startup , He/ Standh)*, Hot Shutdown SSS Inabi li ty to shutdown the reactor causing a cha ll enge to RPV water level or RCS heat removal. Op. Modes: Power Operation 115 ALERT UNUSUAL EVENT SAl Loss of all but one SUl Loss of all offsite AC power source to emergene;' essential buses for 15 minutes or l onger. Op. Modes: Power Operation , Startup , H&I Sumdhy , Hot Shutdown SA2 UNPLANNED l oss of Control Room indications for 15 minutes or l onger w i th a significant transient in progress.
Op. Modes: Power Operation , Startup , He/ Standby , Hot Shutdown SAS Automatic or manual scram fails to shutdown the reactor, and subsequent manual actions taken at the reaetor eontrol eonsoles main control boards are not successful in shutting down the reactor. Op. Modes: Power Operation AC power capabi lit y to emergeney essential buses for 15 minutes or longer. Op. Modes: Power Operation, Startup, H&I Standby , Hot Shutdown SU2 UNPLANNED lo ss of Control Room indications for 15 minutes or longer. Op. Modes: Power Operation , Startup , H&I Sfemdby , Hot Shutdown SU3 Reactor coo l ant activity greater than Technical Specification allowable limits. Op. Modes: Power Operation, Startup , He-I Standby , Hot Shutdown SU4 RCS leakage for 15 minutes or longer. Op. Modes: Power Operation , Startup, He-I Hot Shutdown SUS Automatic or manua l scram fai I s to shutdown the reactor. Op. Modes: Power Operation GENERAL SITE AREA ALERT UNUSUAL EVENT EMERGENCY EMERGENCY SU6 Loss of all onsite or offs i te communications capabilities.
Op. Modes: Pow e r Operation , Startup , &I Skmdhy*, Hot Shutdown SGS Loss of all AC SSS Loss of all Vital and Vital DC power DC power for 15 minutes sources for 15 minutes or or longer. longer. Op. Modes: Power Op. Modes: Power Operation , Startup , &I Operation , Startup , &I &andby*, Hot Shutdown Sffindhy*, Hot Shutdown SA9 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Op. Modes: Power Operation , Startup , &I :; . Hot Shutdown 116

ECL: General Emergency Initiating Conditio n: Prolonged loss of all offsite and all onsite AC power to emergency essential buses. Operating Mode Applicability
Power Operation , Startup , Hot Standby , Hot Shutdown Emergency Action Levels: SG1 Note: The Emergency Director should declare the General Emergency promptly upon determining that (site specific 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months j has been exceeded , or will likely be exceeded.
SG 1.1 a. Loss of ALL offsite and ALL onsite AC power to essential buses 15 and 16specific emergency buses). Basis: AND b. EITHER of the following:
Restoration of at least one essential bus in less than specific 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) is not likely.
Reactor vessel water level cannot be restored and maintained above -149" (Minimum Steam Cooling RPV Water Level) This IC addresses a prolonged loss of all power sources to AC emergency buses essential buses. A loss of all AC power compromises the performance of all SAFETY SYST E MS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control , spent fuel heat removal and the u l timate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers.
In addition , fission product barrier monitoring capabilities may be degraded under these conditions.
The EAL should require declarat i on of a General Emergency prior to meeting the thresholds for IC FG 1. This will allow additional time for imp l ementation of off site protective actions. Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at l east one AC emergency essential bus by the end of the analyzed station blackout coping period. Beyond this time , plant responses and event trajectory are subject to greater uncertainty , and there is an increased likelihood of challenges to multiple fission product barriers.
The estimate for restoring at least one emergency essential bus should be based on a rea l istic appraisa l of the situation.
Mitigation actions with a low probability of success should not be used as a basis for de l aying a classification upgrade. The goal is to maximize the time available to prepare for , and implement , protective actions for the pub l ic. The EAL will also require a Genera l Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core. MNGP Basis Reference(s):
117
1. Ops Man B.09.06-01 (4.16 KV STATION AUXILIARY
-FUNCTION AND GENERAL DESCRlPTION OF SYSTEM) 2. Ops Man C.4-B.09.02.A (STATION BLACKOUT)
3. Ops Man C.4-B.09.02.B (LOSS OF NORMAL OFF-SITE POWER) 4. Ops Man C.4-B.09.06.C (LOSS OF BUS 15 OR BUS 16) 5. C.5.1-1100 (RPV CONTROL) 6. USAR Section 8.2.1 , Plant Electrical Systems -Transmission System , Network Interconnections
7. USAR Section 8.5.1.1, Plant Electrical Systems -DC Power Supply Systems , Essential 250 V de Systems, Design Basis 8. U SAR Section 8.12 , Plant Electrical Systems -Station Blackout 9. USAR Figure 8.4-1, Diesel Generation System One Line Diagram 10. NF-36175 , Single Line Diagram -Station Connection
11. Tech Spec 3.8.1 (AC SOURCES -OPERATING)
12. Tech Spec 3.8.7 (DISTRIBUTION SYSTEMS-OPERATING) 118 SGS ECL: General Emergency Initiating Condition:
Loss of all AC and Vital DC power sources for 15 minutes or longer. Operating Mode Applicability:
Power Operation , Startup , Hot Standb)', Hot Shutdown Emergency Action Levels: Note: The Emergency Director should declare the General Emergency promptl y upon determining that 15 minutes has been exceeded , or will likely be exceeded. SG8. l a. Loss of ALL offsite and ALL onsite AC power to essential buses 15 and 16speeifie emergeney buses) for 15 minutes or longer. Basis: AND b. Indicated voltage is less than 110 VDC (site specific bus 'loltage value) on ALL 125 VDC Vital DC buses (site specific Vital DC busses) for 15 minutes or longer. This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling , containment heat removal/pressure control , spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.
The indicated voltage used in this threshold is based on battery sizing calculations.
The threshold is an average for both Division I and fl batteries for battery voltages at 15 minutes prior to reaching the minimum required terminal voltage. The Division I and ll -250 VDC battery systems need not be considered in this EAL because they supply power to large motor loads in the RCIC and HPCI systems and various non-critical loads. RCTC is an alternative source of make-up water for the reactor during normal plant shutdowns and transient events which lead to a loss of feedwater flow. HPCl is part of the Emergency Core Cooling System (ECCS) network. However , the Auto Depressurization System (ADS) is redundant in function to the HPCI system and does not require 250 VDC for operations.
Therefore , these systems need not be included in this EAL since loss of the 250 VDC battery systems would not cause core uncovering or loss of containment integrity.
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. MNGP Basis Reference(s):
I. Ops Man B.09.06-0l (4.16 KV STATION AUXILIARY
-FUNCTION AND GENERAL DESCRIPTION OF SYSTEM) 2. Ops Man C.4-B.09.02.A (STATION BLACKOUT)
3. Ops Man C.4-B.09.02.B (LOSS OF NORMAL OFF-SITE POWER) 119
4. Ops Man C.4-B.09.06.C (LOSS OF BUS 15 OR BUS 16) 5. USAR Section 8.2. l , Plant Electrical Systems -Tra nsmission System , Network Interconnections
6. USAR Section 8.5.1.1 , Plant Electrical Systems -DC Power Supply Systems , Essential 250 Vdc System, Design Basis 7. USAR Section 8.12 , Plant Electrical Systems -Station Blackout 8. USAR Figure 8.4-1, Diesel Generation System One Line Diagram 9. NF-36175, Single Line Dia gra m -Station Connection I 0. Tech Spec 3.8.1 (AC SOURCES -OPERA TING) 11. Tech Spec 3.8.4 (DC SOURCES -OPERA TING) 12. Tech Spec 3.8.7 (DISTRIBUTION SYSTEMS -OPERA TING) 13. MNGP Calculation CA-02-179, 125 Volt Div. I Calculation
14. MNGP Calculation CA-02-192 , 125 Volt Div. 11 Calculation 120 551 ECL: Site Area Emergency Initiating Condition:
Lo ss of a ll offsite and all onsite AC power to emergency esse nti a l buses for 15 minutes or l onger. Operating Mode Applicability:
Power Operation , Startup , Hot gtandby , Hot Shutdown Emergency Action Levels: Note: The Emergency Director should declare the Site Area E mergenc y promptly upon determining that 15 minutes has been exceeded , or will likel y be exceeded.
SS 1.1 Loss of ALL offsite and ALL onsite AC power to esse n tia l b u ses 1 5 a n d 1 6specific emergency buses) for 15 minutes or longer. Basis: This IC addresses a total loss of AC power that compromises the performance of al l SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling , containment heat removal/pressure control , spent fuel heat removal and the ultimate heat sink. In addition , fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public. Fifteen minutes was se l ected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via ICs RGI , FGl or SGl. MNGP B asis Reference(s)
1. Ops Ma n B.09.06-0 1 (4.16 KV STATfON AUXILIARY-FUNCTION AND GENERAL D ESC R IPTION OF SYSTEM) 2. O p s Man C.4-B.09.0 2.A (STATI ON BLAC K OUT) 3. Ops Ma n C.4-B.09.02.B (LOSS OF NORMAL OFF-SITE POWER) 4. Ops Man C.4-B.09.06.C (LOSS OF BUS 15 O R BUS 1 6) 5. USA R Section 8.2.1 , Plant E l ectrica l Syste m s -Trans mi ssion Syste m , Network I n te r co nn ectio n s 6. USAR F i gure 8.4-1 , Diese l Ge n erat i on System One Li n e Drawing 7. NF-3 61 75, Si n g le L i ne Diagra m -Stat i o n Co nn ect i on 8. Tec h Spec 3.8.1 (AC SOURCES -OPERA TING) 9. Tec h Spec 3.8.7 (DI STRIBUT I ON SYSTEMS -OPE R AT I NG) 121 SSS ECL: Site Area Emergency Initiating Condition:
Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal. Operating Mode Applicability:
Power Operation Emergency Action Levels: SS5. l a. An automatic or manual scram did not reduce reactor power to less than 4%. Basis: AND b. All manual actions to shutdown the reactor are not successful in reducing reactor power to less than 4%. AND c. EITHER of the following conditions exist:
Reactor vessel water level cannot be restored and maintained above -149 (Minimum Steam Cooling RPV Water Level)
Heat Capacity Limit (HCL) exceeded This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown , all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.
In some instances , the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency 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 RG 1 or FG 1. MNGP Basis Reference(s):
1. C.5.1-1100 (RPV CONTROL) 2. C.5.1-1200 (PRIMARY CONTAINMENT CONTROL) 3. C.5.l-2007 (FAILURE TO SCRAM) 122
4. C.5-3101 (ALTERNATE ROD INSERTION)
5. C.4-A (REACTOR SCRAM) 6. USAR Table 7.6-1 , Typical Reactor Protection System Scram Setpoints
7. Tech Spec Table 3.3.1.1-1 (REACTOR PROTECTlON SYSTEM INSTRUMENTATION) 1 23 ECL: Site Area Emergency Initiating Condition:
Los s of all Vital DC power for 15 minutes or longer. Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdown Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergency promptl y upon determining that 15 minutes has been exceeded , or will likely be exceeded.
SSS SS8. l Indicated voltage is less than 110 VDC (site specific bus voltage value) on ALL 125 VDC Vital DC buses (site specific Vital DC busses) for 15 minutes or longer. Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public. The indicated voltage used in this threshold is based on battery sizing calculations. The threshold is an average for both Division I and II batteries for battery voltages at 15 minutes prior to reaching the minimum required terminal voltage. The Division I and II -250 VDC battery systems need not be considered in this EAL because they supply power to large motor loads in the RCIC and HPCI systems and various non-critical loads. RCJC is an alternative source of make-up water for the reactor during normal plant shutdowns and transient events which lead to a l oss of feedwater flow. HPCI is part of the Emergency Core Cooling System (ECCS) network. However , the Auto Depressurization System (ADS) is redundant in function to the HPCI system and does not require 250 VDC for operations.
Therefore , these systems need not be i ncluded in this EAL since loss of the 250 VDC battery systems would not cause core uncovering or loss of containment integrity.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via I Cs RG 1 , FG 1 or SGS. MNGP Basis Reference(s):
1. USAR Section 8.5.1 , Plant Electrical Systems -DC Power Supply Systems , Essential 250 Vdc System 2. USAR Section 8.5.2 , Plant Electrical Systems -DC Power Supply Systems , 125 Vdc System 3. NE-36640-2 , 125VDC Distribution Electrica l Scheme 4. MNGP Calculation CA-02-179, 125 Volt Div. I Calculation
5. MNGP Calculation CA-02-192 , 125 Volt Div. II Calculation 124
6. Tech Spec 3.8.4 (DC SOURCES -OPERA TING) 1 2 5 SA1 ECL: Alert Initiating Condition:
Loss of all but one AC power source to emergency essential buses for 15 minutes or longer. Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdown Emergency Action Levels: Note: The Emergency Director should declare the Alert promptl y upon determining that 15 minutes has been exceeded , or will likely be exceeded.
SAl.1 Basis: a. AC power capability to essential buses 15 and 16 (site specific emergency buses) is reduced to a single power source (Table SI) for 15 minutes or longer. AND b. Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS. Table Sl 1 R Reserve Transformer I AR Reserve Transformer 2R Auxiliary Transformer
I I Emergency Diesel Generator
12 Emergency Diesel Generator SAFETY SYSTEM: A system required for safe plant operation , cooling down the plant and/or placing it in the cold shutdown condition , including the ECCS. These are typically systems classified as safety-related. This IC describes a significant degradation of offsite and onsite AC power sources (Table S 1) such that any additional single failure would resu l t in a loss of all AC power to SAFETY SYSTEMS. In this condition , the sole AC powe r source may be powering one , or more than one , train of safety-related equipment.
This IC provides an escalation path from IC SUI. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supp l ying required power to an emergency essential bus. Some examples of this condition are presented below.
A loss of all offsite power with a concurrent failure of al l but one emergenc y power source (e.g., an onsite diesel generator).
A loss of all offsite pov,'er and loss of all emergenc)' po'Ner sources (e.g., onsite diesel generators) with a single train of emergency busesessential buses being back fed from the unit main generator
A l oss of emergency power sources (e.g., on s ite diesel generators) with a single train of emergency buses essential buses being back-fed from an offsite power source. 126 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 SSL MNGP Basis Reference(s):
1. Ops Man B.09.06-01 (4.16 KV STATTON AUXILIARY-FUNCTION AND GENERAL DESCRIPTION OF SYSTEM) 2. Ops Man C.4-B.09.02.A (STATION BLACKOUT)
3. Ops Man C.4-B.09.02.B (LOSS OF NORMAL OFF-SITE POWER) 4. Ops Man C.4-B.09.06.C (LOSS OF BUS 15 OR BUS 16) 5. USAR Section 8.2. l, Plant Electrical Systems-Transmission System , Network Interconnections
6. USAR Figure 8.4-1, Diesel Generation System One Line Diagram 7. NF-36175 , Single Line Diagram -Station Connection
8. Tech Spec 3.8. l (AC SOURCES -OPERA TING) 9. Tech Spec 3.8.7 (DISTRfBUTION SYSTEMS -OPERA TING) 127 SA2 ECL: Alert Initiating Condition:
UNPLANNED Joss of Control Room indications for 15 minutes or longer with a significant transient in progress.
Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdown Emergency Action Levels: Note: The Emergency Director should dec l are the Alert promptly upon determining that 15 minutes has been exceeded , or will likely be exceeded.
SA2.l a. AND An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer. Reactor Power RPV Water Level RPV Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature
b. ANY of the following transient events in progress.
Automatic or manual runback greater than 25% thermal reactor power
Electrical load rejection greater than 25% full electrical load
Reactor scram
ECCS fSB-actuation
Thermal power oscillations greater than 10% Basis: UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition , the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant. As used in this EAL , an " inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example , the reactor power level cannot be determined from any analog , digital and recorder source within the Control Room. An event involving a loss of plant indications , annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an 128 NRC event report is required.
The event wo uld be reported if it significa ntly impaired the capability to perform emergency assessments.
In particular , emergency assessments necessary to implement abnormal operating procedures , emergency operating procedures , and emergency plan implementing procedures addressing emergency classification, accident assessment , or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control , RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition , if a ll indicati on so urce s 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 va lue for RPV water level cannot be determined from the indications and recorder s on a main control board , the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary l osses of indication.
Escalation of the emergency classification level would be via I Cs FS 1 or IC RS 1. MNGP Basis Reference(s)
1. Ops Man C.4-B.05.13.A (LOSS OF ANNUNCIATOR)
2. Ops Man B.05. l 0 (PROCESS COMPUTER)
3. USAR Section 7.1 , Plant Instrumentation and Control Systems-Summary Description
4. USAR Section 7.13 , Plant I n strumentation and Control Systems -Safety Parameter D i splay System 129 SAS ECL: Alert Initiating Condition:
Automatic or manual scram fails to shutdown the reactor , and subsequent manual actions taken at the reaetor eontro l conso l es main control boards are not successful in shutting down the reactor. Operating Mode Applicability:
Power Operation Note: A manual action is any operator action, or set of actions , which causes the control rods to be rapidly inserted into the core , and does not inc lud e manually driving in control rods or implementation of boron injection strategies. Emergency Action Levels: SAS. I a. An automatic or manual scram did not reduce reactor power to less than 4%. Basis: AND b. Manual actions taken at the reactor contro l conso l es main control boards are not successful in reducing reactor power to less than 4%. This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reaetor co n trol eonso l es main control boards (C-OS) to shutdown the reactor are also unsuccessful.
This condition represents an actua l or potential substantial degradation of the level of safety of the plant. An emergency declaration i s required even if the reactor is subsequentl y shutdown by an action taken away from the reactor contro l consoles main control boards since this event entails a s i gnificant fai lur e of the RPS. A manual action at the reaetor eontro l conso l es main control boards i s any operator action , or set of actions , which causes the contro l rods to be rapidly inserted into the core (e.g., in i tiat i ng a manual reaetor serarn see SUS). This action does not include manually driving in control rods or implementation of boron injection strategies.
If this action(s) i s unsuccessful , operators would immediately pursue additional manual actions at locations away from the reactor contro l conso l es main control boards (e.g., locally openin g breakers).
Actions taken at back-panels or other location s within the Control Room , or any l ocation outside the Contro l Room , are not considered to be " at the reactor contro l conso l es main control boards". Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. [BWR] The plant response to the failure of an automatic or manual reactor scram will vary based upon severa l factors including the reactor power level prior to the event , avai l ability o f the condenser , performance of mitigation equipment and act ion s , other conc urr ent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the RPV water level or RCS heat removal safety functions , the emergency classification level will escalate to a Site Area Emergency via IC SSS. Depending upon plant responses and symptoms , escalation is als o 130 possible via IC FS 1. Absent the plant conditions needed to meet either IC SSS or FS 1 , an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria. The MNGP EOP entry condition for a failure to scram is defined to be a power above the APRM downscale setpoint (4%) following a reactor scram indicates that significant power is being generated.
lf a scram is successful , reactor power will be indicated to be less than 4%. Therefore , reducing power to LESS THAN 4% is used in SAS. I as indication that a scram was successful.
MNGP Basis Reference(s):
I. C.4-A (REACTOR SCRAM) 2. C.S. l-1000 (EOP INTRODUCTION)
3. C.S.l-1100 (RPV CONTROL-FLOWCHART)
4. C.S.l-2007 (FAILURE TO SCRAM) S. C.S-3101 (ALTERNATE ROD INSERTION)
6. USAR Table 7.6-1, Typical Reactor Protection System Scram Setpoints
7. Tech Spec Table 3.3.1.1-l (REACTOR PROTECTION SYSTEM INSTRUMENTATION) 131 SA9 ECL: Ale1i Initiating Condition:
Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdown Emergency Action Levels: SA9.l a. The occurrence of ANY of the following hazardous events: Basis: AND
Seismic event (Earthquake)
Internal or externa l flooding event
High winds or tornado strike
FIRE
EXPLOSION
River level greater than 919 ft el.
River level less than 900.5 ft el.
Other events with similar hazard characteristics as determined by the Shift Manager b. EITHER of the following:
Event damage has caused indications of degraded performance in at least one tra in of a SAFETY SYSTEM needed for the current operating mode.
The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode. EXPLOSION:
A rapid , violent and catastrophic failure of a piece of equipment due to combustion , chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits , grounding , arcing , etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present. FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
SAFETY SYSTEM: A system required for safe plant operation , cooling down the plant and/or placing it in the cold shutdown condition , including the ECCS. These are typically system s classified as safety-related. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements , testing , or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM , or a structure contain in g SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier , 132 and therefore represents an actual or potential substantial degradation of the level of safety of the plant. The first threshold for EAL addresses damage to a SA F ETY SYSTEM train that is in service/operation since indications for it will be readily available.
The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. The second threshold for EAL-h SA9. l damage to a SAFETY SYST E M component that is not in service/operation or readily apparent through indications alone , or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC FS 1 or RS 1. MNGP Basis Reference(s):
1. Ops Man A.6 (ACTS OF NATURE) 2. Ops Man B.05.14 (SEISMIC MONITORING SYST E M) 3. Ops Man B.05.16-01 (METEOROLOGICAL MONITORING
-F UNCTION & GENERAL DESCRIPTION OF SYSTEM) 4. Ops Man B.06.04 (CIRCULATING WATER SYST E M) 5. C.4-B.05.14.A (EARTHQUAKE)
6. C.6-006-C-08 (EARTHQUAKE)
7. C.6-006-C-13 (OPERA TI ON AL BASIS EARTHQUAKE)
8. USAR Section 10.3 , Plant Auxiliary Systems -Plant Service Sy s tems 9. USAR Section 12.2 , Plant Structures and Shieldin g, Plant Principal Structur es and Foundations
10. USAR Appendix G , Chapter 3 , Probable Maximum Flood Determination
11. USAR Table I.5-1 , Location of High Energy Systems and Safe Shutdown Equipment by Volume 12. USAR Appendix J.4 , Fire Protection Program -Safe Shutdown Analysis 13. ND-95208 , Monticello Property Map 14. ND-95209 , Monticello Main Plant Structur e s 15. 4 AWI-01.03.01 (QUALITY ASSURANCE PROGRAM BOUNDARY) 133 SU1 ECL: Notification of Unusual Event Initiating Condition:
Loss of all offsite AC power capability to emergeney essential buses for 15 minutes or longer. Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdown Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded , or will likely be exceeded. SUI .1 Loss of ALL offsite AC power capability (Table S2) to essential buses 15 and 16speeifie emergeney buses) for 15 minutes or longer. Table S2 1 R Reserve Transformer 1 AR Reserve Transformer 2R Auxiliary Transformer Basis: This IC addresses a prolonged loss of offsite power (Table S2). The loss of offsite power sources renders the plant more vu lnerable to a complete loss of power to AC emergeney buses. This condition represents a potential reduction in the level of safety of the plant. For emergency classification purposes , " capabi lit y" means that an offsite AC power source(s) is available to the emergeney buses essential buses , whether or not the buses are powered from it. Fifteen minutes was se l ected as a threshold to exclude transient or momentary losses of offsite power. Escalation of the emergency classification level would be via IC SAl. MNGP Basis Reference(s):
1. Ops Man B.09.06-01 (4.16 KV STATION AUXILIARY-FUNCTION AND GENERAL DESCRIPTION OF SYSTEM) 2. Ops Man C.4-B.09.02.A (STATION BLACKOUT)
3. Ops Man C.4-B.09.02.B (LOSS OF NORMAL OFF-SITE POWER) 4. Ops Man C.4-B.09.06.C (LOSS OF BUS 15 OR BUS 16) 5. USAR Section 8.2.1 , Plant Electrical Systems -Transmission System , Network Interconnections 134
6. USAR Figure 8.4-1, Diesel Generation System One Line Diagram 7. NF-36175 , Single Line Diagram -Station Connection
8. Tech Spec 3.8.1 (AC SOURCES -OPERATING)
9. Tech Spec 3.8.7 (DISTRIBUTION SYSTEMS -OPERA TING) 135 SU2 ECL: Notification of Unusual Event Initiating Condition:
UNPLANNED loss of Control Room indications for 15 minutes or longer. Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdown Emerge nc y Action Levels: Note: The Emergency Director should declare the Unusua l Event prompt l y upon determining that 15 minutes has been exceeded, or will likely be exceeded. SU2.l Basis: An UNPLANNED event results in the inability to monitor one or more of the follow* t fl "th" th C t 1 Room for 15 minutes or longer. mg parame ers rom wt m e on ro Reactor Power RPV Water Leve l RPV Pressure Primary Containment Press ur e Suppression Poo l Level Suppression Pool Temperature UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the l eve l of safety of the plant. As used in this EAL , an " inability to monitor" means that va lu es for one or more of the li sted parameters cannot be determined from within the Control Room. This s itu at i on wou ld require a loss of a ll of the Contro l Room sources for the given parameter(s).
For example , the reactor power l evel cannot be determined from any ana l og , digital and recorder source within the Control Room. An event involving a loss of plant indications , annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-10 22) to determine if an NRC event report is required.
T h e event wo uld be reported i f it significant l y impaired the capability to perform emergency assessments. In particular , emergency assessments necessary to implement abnormal operating procedures , emergency operating procedures , and emergency plan implementing procedures addressing emergency class i ficat i on , 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 , RPV level and RCS heat removal. The l oss of the ability to determine one or more of these parameters from within the Control Room is considered to be 136 more significant than simply a reportable condition.
In addition , if all indication sources for one or more of the listed parameters are Jost , then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example , ifthe value for RPV water level cannot be determined from the indications and recorders on a main control board , the SPDS or the plant computer , the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level would be via IC SA2. MNGP Basis Reference(s):
1. Ops Man C.4-B.05.13.A (LOSS OF ANNUNCIATOR)
2. Ops Man B.05.10 (PROCESS COMPUTER)
3. USAR Section 7.1 , Plant Instrumentation and Control Systems -Summary De s cription 4. USAR Section 7.13 , Plant Instrumentation and Control Systems -Safety Parameter Display System 137 SU3 ECL: Notification of Unusual Event Initiating Condition:
Reactor coolant activity greater than Technical Specification allowable limits. Operating Mode Applicability:
Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (SU3.1 or SU3.2) SU3.l Offgas Pretreatment Radiation Monitor (RM-l 7-150A or RM-l 7-150B) high radiation alann (4-A-12) received (Site specific radiation monitor) reading greater than (site specific value). SU3.2 Coolant sample activity greater than 0.2 µCi/gm dose equivalent 1-131.Sample analysis indicates that a reactor coolant activity value is greater than an allo*Nable limit specified iA Technical Specifications. Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. For EAL SU3.1 the Offgas high radiation alarm received from RM-17-150A or RM-17-1508 is set to meet the Technical Specification allowable limit of less than or equal to 2.6E+5 µCi/sec after decay of 30 minutes. EAL SU3.2 addresses reactor coolant samples exceeding coolant Technical Specifications.
Escalation of the emergency classification level would be via ICs FAl or the Recognition Category R ICs. MNGP Basis Reference(s):
1. Ops Man B.07.02.02-01 (OFF-GAS HOLDUP SYST E M-FUNCTION & G E NERAL DESCRIPTION OF SYSTEM) 2. C.6-004-A-12 (OFF-GAS Hl RADIATION)
3. Tech Spec 3.4.6 (RCS SPECIFIC ACTIVITY)
4. Tech Spec 3.7.6 (MAIN CONDENSER OFFGAS) 5. Tech Spec 3.10.1 (INSERVTCE LEAK AND HYDROSTATIC TESTING OPERATION) 138 SU4 ECL: Notification of Unusual Event Initiating Condition:
RCS leakage for 15 minutes or longer. Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdown Emergency Action Levels: (SU4.l or SU4.2 or SU4.3) Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded , or will likely be exceeded.
SU4.l SU4.2 SU4.3 Basis: RCS unidentified or pressure boundary leakage greater than (site specific va l ue) 10 gpm for 15 minutes or longer. RCS identified leakage greater than (s i te s pecific va l ue)25 gpm for 15 minutes or longer. Leakage from the RCS to a location outside primary containment greater than 25 gpm for 15 minutes or longer. This IC addresses RCS leakage which may be a precursor to a more significant event. In this case , RCS leakage has been detected and operators , following applicable procedures , have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safet y of the plant. EAL #SU4. l and EAL #SU4.2 are focused on a loss of mass from the RCS due to "unidentified leakage" , "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications).
EAL #SU4.3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs thus apply to leakage into the containment or a location outside of containment.
The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications.
Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). EAL #SU4. l uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage. The release of mass from the RCS due to the as-designed
/expected operation of a relief valve does not warrant an emergency classification.
A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and , therefore , is not applicable to this EAL. 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 I Cs of Recognition Category R or F. 139

MNGP Basis Reference(s):
1. Ops Man C.4-B.04.01.F (LEAK INSIDE PRIMARY CONTAINMENT)
2. Ops Man C.6-004-B-03 (DR YWELL SUMP VAL YES CLOSED) 3. Ops Man C.6-004-B-17 (DRYWELL FLOOR DRAIN SUMP HI LEVEL) 4. Ops Man C.6-004-B-18 (DRYWELL EQUIP DRAIN LEAK RATE CHANGE HI) 5. Tech Spec 3.4.4 (RCS OPERATIONAL LEAKAGE) 140 SUS ECL: Notification of Unusual Event Initiating Condition:
Automatic or manual scram fails to shutdown the reactor. Operating Mode Applicability:
Power Operation Note: A manual action is any operator action , or set of actions , which causes the control rods to be rapidly inserted into the core , and does not include manually driving in control rods or implementation of boron injection strategies. Emergency Action Levels: (SUS.I or 8U5.2) SUS.I a. An initial automatic or manual scram did not reduce reactor power to less than 4%. AND b. A subsequent manual action taken at the reactor control consoles main control 8U5.2 a. Basis: AND b. boards is successful in shutting down the reactor. A manual scram did not shutdown the reactor. EITHER ANY of the following is successful in reducing reactor power to less than 4%:
Manual scram pushbuttons
Mode switch to shutdown
Alternate rod insertion (ARI)
Subsequent automatic s cram This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown , and either a subsequent operator manual action taken at the reactor control eonsoles main control boards (C-OS) or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Following the failure on an automatic reactor scram , operators will promptly initiate manual actions at the reactor control consoles main control boards to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor , core heat generation wi ll quickly fall to a l evel within the capabilities of the plant's decay heat removal systems. If an initial manual reactor scram is unsuccessful , operators will promptly take manual action at another location (s) on the reactor control consoles main control boards to shutdown the reactor (e.g., initiate a manual reactor seram manual scram pushbuttons , mode switch to Shutdown, Alternate Rod Insertion (ARI)) using a different switch). Depending upon several factors , the initial or subsequent effort to manually scram the reactor , or a concurrent plant condition, may lead to the generation of an automat ic reactor scram signal. If a subsequent manual or automatic 141 scram is successful in shutting down the reactor , core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the reactor control consoles main control boards is any operator action , or set of actions , which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies.
Actions taken at back-panels or other locations within the Control Room , or any location outside the Control Room , are not considered to be "at the reactor control consoles main co n tro l boards". Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. [BWR] The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event , availability of the condenser , performance of mitigation equipment and actions , other concurrent plant conditions , etc. If subsequent operator manual actions taken at the reactor control consoles ma i n control boa r ds are also unsuccessful in shutting down the reactor , then the emergency classification level will escalate to an Alert via IC SAS. Depending upon the plant response , escalation is also possible via IC FAl. Absent the plant conditions needed to meet either IC SAS or FAl , 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 scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.
If the signal causes a plant transient that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor , then this IC and the E ALs are applicable , and should be evaluated.
If the signal does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.
MNGP Basis Reference(s):
1. C.4-A (REACTOR SCRAM) 2. C.S.1-1000 (EOP INTRODUCTION)
3. C.S.1-1 1 00 (RPV CONTROL -FLOWCHART)
4. C.S.1-200 7 (FAILURE TO SCRAM) S. C.S-3 1 0 1 (ALTERNATE RO D IN SERT I ON) 6. USAR Table 7.6-1 , Typ i ca l Reactor Protectio n System Scram Setpo i nts 7. Tec h Spec Ta bl e 3.3.1.1-1 (R EACTOR PROTECT I ON SYSTEM INSTRUMENTATION) 142 SU6 ECL: Notification of Unusual Event Initiating Condition:
Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdown Emergency Action Levels: (SU6.l or SU6.2 or SU6.3) SU6.l Loss of ALL of the following onsite communication methods:
Commercial Telephones
Plant Telephones
Portable radios
Plant PA System (site speeific list of communications methods) SU6.2 Loss of ALL of the following Offsite Response Organization (ORO) communications methods: (site specific list of communications methods)
Commercial Telephones
Direct Dedicated Telephones
Radio/Receiver Transmitter SU6.3 Loss of ALL of the following NRC communications methods: (site specific I ist of communications methods)
Federal Telecommunications System (FTS)
Commercial Telephones Basis: This IC addresses a significant loss of on-site or off site communications capabilities.
While not a direct challenge to plant or personnel safety , this event warrants prompt notifications to OROs and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment , relaying of on-site information via individuals or multiple radio transmission points , individuals being sent to offsite locations , etc.). EAL #S U6. l addresses a total loss of the communications methods used in support of routine plant operations.
EAL #SU6.2 addresses a total loss of the communications methods used to notify all OR Os of an emergency declaration.
The OR Os referred to here are (see Developer Notes) the State of Minnesota , Wright County , and Sherburne County. EAL #SU6.3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
143 MNGP Basis Reference(s):
1. USAR Section 10.3.8 , Plant Auxiliary Systems -Plant Service Systems , Plant Communications System 2. MNGP Emergency Plan Section 7.2 -Communication Systems 3. MNGP Emergency Plan Figure 13.7 -Direct Dedicated Telephones (Hot Lines) 144 APPENDIX A -ACRONYMS AND ABBREVIATIONS AC ............
................
.............
.............................................................................
Alternating Current AOP ..........................
.............................
............................
..............
Abnormal Operating Procedure APRM ..................................
................................
...............................
Average Power Range Meter A RI ..........................................
................
....................
.......................
.......... A lt e rn ate R o d In sert i o n A TWS ...................................................................................
Anticipated Trans i ent Witho u t Scram ..............................
...............................................................
.............................................
B &W .....................
.......................
..................................................
..................... Babcock and Wilco>< .......................................................................................................................................... B HT ..............
.........................................................................
Boron Injection Initiation Temperature BWR .......................................................................
..............
........................ Boiling Water Reactor CDE ........................................
...........................................................
... Committed Dose Equ i valent CFR ...................
...................................................................................
Code of Federal Regulations CTMT /CNMT ....................
..........................................
................................................. Containment
..................
...............................
.............................................
............................................ c SF .........................
............................................................................
........... Critical Safety Function ..........................................................................................................................................
c SFST ..............
............................................
..............................
Critical Safety function Status Tree DBA .............................
.................................................................................
De s ign Basis Accident DB E ............
............................................................................................... D es i g n Bas i s Ea rthqu ake DC ...................
...........................
..........
.............................................
......................... Direct Current D SC ..............................................
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........................... Dry S h i e l d e d Cas k EAL ........................
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.................... Emergency Action Leve l ECCS .................................
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Emergency Core Coo lin g System ECL .......................................
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...... Eme r ge ncy Classification Leve l EFT ..........
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................................................... Ex h a u st F il tr a ti o n T r a in EOF ....................
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.......... Eme rgency Operations Facility EOP .........................
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...... Emergency Operating Procedure EPA ..................................................
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... Enviro nmental Protection Agency EPG ............
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Emerge nc y Procedure Guideline EPIP ..............................
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................................
Emergency Plan Implementing Procedure
..................
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... E PR .........................................................................................
................
E!.volutionary Po\ver Reactor EP RI ........................................
...............................................
...... E l ectric Power Research Institute ERG ................................................................................................
Eme rgency Re spo nse G ui deline FAA ...........................................................................
.................... Federa l Av i a ti o n A dmin i strat i o n F BI ................
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.................... Fe d era l B u r ea u of I n vest i gat i o n FEMA .....................
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Federa l Emergency Management Agency F PB ...........................................................
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............................... F i ss i o n P ro d u ct Ba rri e r FSAR ....................
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..................
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Fina l Safety Analysis Report GE .....................................................................................
................................. Genera l Emergency HC T L ......................
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........................................................
Heat Capacity Temperatl1re Limit HO O ...............................
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........................................... Hea dquart e r s Ope r a tion O ffice r HPCI ..................
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..................
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...... High Pressure Coo l ant Injection H SM ..........
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........................... H or i zo nt a l Sto r age Mod ul e ......................
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...........
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..................... H SI ..................
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.................... Human System Interface IC .......................
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...............
............................................................
Initiating Condit i on ID ...........
...............................................
........................................
...........................
Inside Diameter A-1 IPEEE ..................
...........
Individual Plant Examination of External Events (Generic Letter 88-20) ISFSI ..............
.........................................................
.... Independent Spent Fuel Storage Installation Keff ............
......................
..................................................
Effective Neutron Multiplication Factor LCO ............
.......................
............................................................
Limiting Condition of Operation LOCA ..............................
.........................................................
................. Loss of Coolant Accident MCR ..................................
...............................
......................
........................... Main Control Room MSIV .....................................................................................................
Main Steam Isolation Valve MSL ...........
...........
.....................................................................................
............ Main Steam Line mR , mRem , mrem, mREM ..........................
..........................
........ mi Iii-Roentgen Equivalent Man MW .................................................................................
................................................... Megawatt NEI ............................................................
.............................
.................... Nuclear Energy Institute NORAD ...........................................................
...... North American Aerospace Defense Command NPP ............
.......................
........................................................................
....... Nuclear Power Plant NRC ..........................................................
.................................... Nuclear Regulatory Commission NSPM ..........
..........
..........................
.... Northern S t ates Power Company, a M inn esota cor p oration, ......................................................
..............
...................................... do i ng bu s in ess as Xce l Energy NSSS .................................................................................................
Nuclear Steam Supply System .......................................................................................
.... (N otification O f) Unusual Event ....................................
...............
.....................................
..............................................
.... t'J" UMARC 6 ........................................
.................
......... Nuclear Management and Resources Council OBE .....................................................................
..................................
Operating Basis Earthquake OCA ......................
.....................
..................................................................
Owner Controlled Area ODCM/ODAM ..........................
........................
.... Offsite Dose Calculation (Assessment)
Manual ORO ...................
........................
.............
........................................ Off-site Response Organization PA .........................
....................
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............................................................. Protected Area ................................................................................................
.......................................... P AC8 ......................................................................................
Priority Actuation and Control System PAG ...............
....................
........................
..........................
.................. Protective Action Guideline PCIS ............
...................
................................................
..... Pri m ary Conta in ment Iso l a ti on System ..........................................................
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..........................................
p 1C8 ............................
.........................
..............................
Process Information and Control System PRA/PSA ............
........................
Probabilistic Risk Assessment I Probabilistic Safety Assessment
......................
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.... : ..............
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p \VR ..........................................................................................
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Pr ess uri ze d \Voter Reactor ................................................................
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p 8 .................
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...............
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Protection 8yste1n PSIG ....................
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............ : .......................... Pounds per Square Inch Gauge R ...........................
...............................
...........................
.......................
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Roentgen RCC ................................
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................
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............. Reactor Control Console RCIC ................................
...............................................................
Reactor Core Isolation Cooling RCS ..........
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..............................
...................................
..................... Reactor Coolant System Rem , rem, REM ....................................
..................................................
Roentgen Equivalent Man RETS ............................................
..............
............. Radiological Effluent Technical Specifications RHR ...............................
............................
.............................
..................... Res i dua l f-lea t Remova l RPS ..................................................
...............
........................................
Reactor Protection System RPV ....................
.....................
............................................................
........ Reactor Pressure Vessel RVLIS .....................................
.................................
Reactor Vessel Level Instrumentation System RWCU ................
..........................................................................................
Reactor Water Cleanup (;NUMA RC was a pree:leeessor organizat i o n of the N1:1elear Energy lnstih1te (NEI). A-2 SAE ................................
.........................
..............................
...............
........... Site Area E1nergency SAMG .................................................
...............
..........
... Severe Acci dent Management Guidelines SAR ..................
..............................................
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...................
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Safety Ana l ysis Report .....................................
..................................................................................................... s 1\S .......................................................................................
................... Safety /\.utomation System SBO .............
....................
...............
................................
................
..................
....... Stat ion Blackout SCBA . . . . . . . . . ...............
.. . . . . ...... .. . . . . . . . . . . . . .. .........................
........ Self-Contained Breathing Apparat u s ****************************************************************************************************************************************** s G ..............
................................................................................................................ Steam Generator
.......................................................................................................................................... s I .......................................................................................................
......................... Safety l njeetion .......................................................................................................................................... s TGS ............................................................
......................... Safet)' Information and Contro l System SPDS ..............
....................
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......................
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Safety Parameter Display System SRO ...................
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.............................
...........................................
Se nior Reac tor Operator SRV ..................
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Safety Relief Valve T AF ....................
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Top of Active Fuel TBNWS .............
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Turbine Building Normal Waste Sump TEDE ...................
........................
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........ Total Effective Dose Eq ui va l ent TOAF ...................................................................
..........................
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........ Top of Aetive Fue l TSC .................................................................
.........................
................ Technical S upp ort Center USAR .............................................................................................
Update d Safety Analysis R epo rt .......................................................................................................................................... \V OG ...................................................................................................... Westinghouse 0\vners Group A-3 APPENDIX B -DEFINITIONS The following definitions are taken from Title 10 , Code of Federal Regulations , and related regulatory guidance documents.
Alert: Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels. General Emergency (GE): Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Notification of Unusual Event (N G UE)+: Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs. Site Area Emergency (SAE): Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could l ead to the likely failure of or; 2) that prevent effective access to , equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.
The following are key terms necessary for overall understanding the NEI 99-01 emergency classification scheme. Emergency Action Level (EAL): A pre-determined , site-specific , observable threshold for an Initiating Condition that , when met or exceeded , places the plant in a given emergency classification level. Emergency Classification Level (ECL): One of a set of names or titles established by the US Nuclear Regulatory Commis s ion (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences , and (2) resulting onsite and off site response actions. The emergency classification levels , in ascending order of severity , are:
Not i fication of Unusual Event (N G UE)
Alert
Site Area Emergency (SAE)
General Emergency (GE) Fission Product Barrier Threshold:
A pre-determined , site-specific , observable thresho ld +This is sometimes shorteAeEI to UAus1:1al B*leAt (lie) or other similar site termiAology.
B-1 indicating the loss or potential loss of a fission product barrier. Initiating Condition (IC): An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.
Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document.
The definitions of these terms are provided below. CONFINEMENT BOUNDARY:
The barrier(s) between areas containing spent fuel and the environment once the spent fuel is processed for dry storage. (Insert a site specific definition for this term.) CONTAINMfillT CLOSURE (Insert a site specific definition for this term.) EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion , chemical reaction or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits , grounding, arcing , etc.) should not automatically be considered an explosion.
Such events may require a post-event inspection to determine if the attributes of an explosion are present. FIRE: Combustion characteri z ed by heat and light. Sources of smoke such as s lipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat a re observed. HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment , take HOSTAGES , and/or intimidate the licensee to achieve an end. Thi s includes attack by air , land , or water using guns; explosives , PROJECTILEs , vehicles , or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts betwe e n individuals in the owner controlled area). HOSTILE FORCE: One or more individuals who are engaged in a determined assault , overtly or by stealth and deception , equipped with suitable weapons capable of killing , maiming , or causing destruction.
IMMINENT:
The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. R-2 ISFSI PROTECTED AREA: The area surrounding the Independent Spent Fuel Storage Installation encompassed by the double chain link fence surrounding the ISFSI as defined in the Security Plan; the ISFSI Protected Area is excluded from the Plant Protected Area. NORMAL LEVELS: As applied to radiological IC/EALs , the highest reading in the past four hours excluding the current peak value. OWNER CONTROLLED AREA: The OCA boundaries consist of the plant property enclosed by a three strand barbed wire fence and a posted boundary on the Wright County side of the nver. (Insert a site specific definition for this term.) PROJECTILE
An object directed toward a NPP that could cause concern for its continued operability , reliability , or personnel safety. PROTECTED AREA: The area surrounding the plant encompassed by the chain link fence and certain structures as defined in the Security Plan; excludes the lSFSl Protected Area. In areas where two fences are present , the inner fence is designated as the Protected Area barrier (Tnsert a site specific definition for this term.) . REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool, or fuel transfer canal.(lnsert a site specific definition for this term.) SAFETY SYSTEM: A system required for safe plant operation , cooling down the plant and/or placing it in the cold shutdown condition , including the ECCS. These are typically systems classified as safety-related. SECURITY CONDITION:
Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel , or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION. SECONDARY CONTAINMENT:
SECONDARY CONTAINMENT includes the Reactor Building (including the HPCI Building), the Standby Gas Treatment System , the Offgas Dilution Fans , and connecting pipes and ducts. SECONDARY CONTAINMENT is isolated along with an automatic initiation of the Standby Gas Treatment System to minimize radiological releases to the environment.
SITE BOUNDARY:
For Dose Assessment and Protective Action Recommendation purposes the SITE BOUNDARY is the closest distance at wh i ch members of the public would be exposed to a radioactive release. The SITE BOUNDARY for l i quid releases of radioactive material is defined in ODCM-02.01 (UQUID EFFLUENTS).
T h e SITE BOUNDARY for gaseous releases of radioactive material is defined in ODCM-03.01 (GASEOUS EFFLUENTS).
UNISOLABLE:
An open or breached system line that cannot be isolated , remotely or locally. B-3 UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
B-4 L-MT-17-012 NSPM ATTACHMENT 3 MONTICELLO NUCLEAR GENERATING PLANT License Amendment Request to Revise the Emergency Action Level Scheme Emergency Action Level Technical Bases Document (Clean Copy Version) ( 146 pages to follow)
MONTICELLO NUCLEAR GENERATING PLANT EMERGENCY ACTION LEVELS INITIATING CONDITIONS, THRESHOLD VALUES, AND BASES TABLE OF CONTENTS 1 REGULATORY BACKGROUND
A************************************
1 1.1 0PERATINGREACTORS
.*************.***.****..********.*.**************
l 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSl) ***********.*************************
l 1.3 NRC ORDER EA-12-051
.***
2 2 MNGP KEY TERMINOLOGY
a****************
4 2.1 EMERGENCY CLASSIFICATION LEVEL (ECL) ****.*.*****...*********.*.****.**********.**************.****
.4 2.2 INITIATING CONDITION (IC) **********************************************.****************************.**************
6 2.3 EMERGENCY ACTION LEVEL (EAL) *****************************************************************************
6 2.4 FISSION PRODUCT BARRIER THRESHOLD

6 3 DESIGN OF THE NEI 99*01 EMERGENCY CLASSIFICATION SCHEME ADOPTED BY MNGP ***************************************************11***********************************************************************
8 3 .1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLs) *******************************
8 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS **********.*****.***
11 3.3 MNGP SPECIFIC DESIGN CONSIDERATIONS
...**********.************.****.*********************
11 3.4 ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION
.***************.***
12 3.5 IC AND EAL MODE APPLICABILITY

12 4 MNGP SCHEME DEVELOPMENT GUIDANCE .......................................**..***............*.
14 4.1 GENERAL IMPLEMENTATION GUIDANCE ********************************************************************
14 4.2 CRITICAL CHARACTERISTICS
.*****************************.**********************.*.*....
14 4.3 INSTRUMENTATION USED FOR EALs ***************************************.********.*************************
15 4.4 PRESENTATION OF SCHEME INFORMATION TO USERS ***********************************.***********
15 4.5 INTEGRATION OF ICs/EALs WITH PLANT PROCEDURES
.****.********************************
16 4.6 BASIS DOCUMENT ***.***.********************************************************************.*****.*****..*********.*....
16 4. 7 EAL/THRESHOLD REFERENCES TO MNGP AOP AND EOP SETPOINTS/CRITERIA
.17 5 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS
......................................
18 5.1 GENERAL CONSIDERATIONS
.***********.***************************************
18 5.2 CLASSIFICATION METHODOLOGY

19 5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS

19 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION
.******..****....
19 5.5 CLASSIFICATION OF IMMINENT CONDITIONS
.***************.********.***.****************************
20 5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING
.*..***
20 5.7 CLASSIFICATION OF SHORT-LIVED EVENTS ...............................................................
20 5.8 CLASSIFICATION OF TRANSIENT CONDITIONS
..*******************************************************
20 5.9 AFTER-THE-FACTDISCOVERYOFANEMERGENCYEVENTORCONDITION
.**...*
21 5 .10 RETRACTION OF AN EMERGENCY DECLARATION
...*******.***************..***.*****.************.*
22 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS ************************
23 7 COLD SHUTDOWN/
REFUELING SYSTEM MALFUNCTION ICS/EALS ********************
41 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS **************
66 9 FISSION PRODUCT BARRIER ICS/EALS ******************************************************************
68 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS *********
83 11 SYSTEM MALFUNCTION ICS/EALS *************************************************************************
109 APPENDIX A -ACRONYMS AND ABBREVIATIONS

A-1 APPENDIX B -DEFINITIONS

8-1 ii DEVELOPMENT OF EMERGENCY ACTION LEVELS FOR MONTICELLO NUCLEAR GENERATING PLANT 1 REGULATORY BACKGROUND 1.1 OPERATING REACTORS Title 10, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities.
Several of these regulations govern various aspects of an emergency classification scheme. A review of the relevant sections listed below will aid the reader in understanding the key terminology provided in Section 3.0 of this document.
10 CFR § 50.47(a)(l)(i)
10 CFR § 50.47(b)(4)
10 CFR § 50.54(q)
10 CFR § 50.72(a)
10 CFR § 50, Appendix E, IV .B, Assessment Actions
10 CFR § 50, Appendix E, IV.C, Activation of Emergency Organization The above regulations are supplemented by various regulatory guidance documents.
Three documents of particular relevance to NEI 99-01 are:
NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants]
NUREG-1022, Event Reporting Guidelines 10 CFR § 50. 72 and§ 50. 73
Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) Selected guidance in NEI 99-01 is applicable to the Monticello Nuclear Generating Plant (MNGP) emergency plan to fulfill the requirements of 10 CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CFR § 50 and the guidance in NUREG 0654/FEMA-REP-l.
The initiating conditions germane to a 10 CFR § 72.32 emergency plan (as described in NUREG-1567) are subsumed within the classification scheme for a 10 CFR § 50.47 emergency plan. The MNGP IC and EAL for an ISFSI are presented in Section 8, ISFSI ICs/EALs.
IC EUl covers the spectrum of credible natural and man-made events included within the scope of the ISFSI design. In addition, appropriate aspects oflC HUl and IC HAl address a HOSTILE ACTION directed against the ISFSI. The analysis of potential onsite and off site consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-1140, A Regulatory Analysis on 1 Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees.
NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 rem Effective Dose Equivalent.
Regarding the above information, the expectations for an offsite response to an Alert classified under a 10 CFR § 72.32 emergency plan are generally consistent with those for a Notification of Unusual Event in a 10 CFR § 50.47 emergency plan (e.g., to provide assistance if requested).
Also, the MNGP Emergency Response Organization (ERO) required for a 10 CFR § 72.32 emergency plan is different than that prescribed for a 10 CFR § 50.47 emergency plan (e.g., no emergency technical support function).
1.3 NRC ORDEREA-12-051 The Fukushima Daiichi accident of March 11, 2011, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity, and ultimately led to core damage in three reactors.
While the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to preclude fuel damage from the loss of cooling. Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii).
Among them was to provide each spent fuel pool with reliable level instrumentation to significantly enhance the ability of key decision-makers to allocate resources effectively following a beyond design basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all US nuclear plants with an operating license, construction permit, or combined construction and operating license. NRC Order EA-12-051 states, in part, "All licensees
... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel:
(1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." To this end, all licensees must provide:
A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool;
A display in an area accessible following a severe event; and
Independent electrical power to each instrument channel and provide an alternate remote power connection capability.
2 _ __J NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", provides guidance for complying with NRC Order EA-12-051.
This document includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-051.
These EALs are included within existing IC RA2, and new I Cs RS2 and RG2. 3 2 MNGP KEY TERMINOLOGY There are several key terms that appear throughout the emergency classification methodology for MNGP. These terms are introduced in this section to support understanding of subsequent material.
As an aid to the reader, the following table is provided as an overview to illustrate the relationship of the terms to each other. Emergency Classification Level GE I SAE I Alert I Unusual Event Initiating Condition Initiating Condition Initiating Condition Initiating Condition Emergency Action Emergency Action Emergency Action Emergency Action Level (1) Level (1) Level (1) Level (1)
Operating Mode
Operating Mode
Operating Mode
Mode Applicability Applicability Applicability Applicability
Notes
Notes
Notes
Notes
Basis
Basis
Basis
Basis (1) -When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition.
This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basis information.
In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL. 2. l EMERGENCY CLASSIFICATION LEVEL (ECL) One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:
Notification of Unusual Event (NUE)
Alert
Site Area Emergency (SAE)
General Emergency (GE) 2.1.1 Notification of Unusual Event (NUE) Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs. 4 Purpose: The purpose*of this classification is to assure that the first step in future response has been carried out, to bring the operations staff to a state of readiness, and to provide systematic handling of unusual event information and decision-making.
2.1.2 Alert Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels. Purpose: The purpose of this classification is to assure that emergency personnel are readily available to respond ifthe situation becomes more serious or to perform confirmatory radiation monitoring if required, and provide off site authorities current information on plant status and parameters.
2.1.3 Site Area Emergency (SAE) Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.
Purpose: The purpose of the Site Area Emergency declaration is to assure that emergency response centers are staffed, to assure that monitoring teams are dispatched, to assure that personnel required for evacuation of near-site areas are at duty stations ifthe situation becomes more serious, to provide consultation with offsite authorities, and to provide updates to the public through government authorities.
2.1.4 General Emergency (GE) Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA P AG exposure levels off site for more than the immediate site area. Purpose: The purpose of the General Emergency declaration is to initiate predetermined protective actions for the public, to provide continuous assessment of information from the licensee and offsite organizational measurements, to initiate additional measures as indicated by actual or potential releases, to provide consultation with offsite authorities, and to provide updates for the public through government authorities.
5 2.2 INITIATING CONDITION (IC) An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.
Discussion:
An IC describes an event or condition, the severity or consequences of which meets the definition of an emergency classification level. An IC can be expressed as a continuous, measurable parameter (e.g., RCS leakage), an event (e.g., an earthquake) or the status of one or more fission product barriers (e.g., loss of the RCS barrier).
Appendix 1 of NUREG-0654 does not contain example Emergency Action Levels (EALs) for each ECL, but rather Initiating Conditions (i.e., plant conditions that indicate that a radiological emergency, or events that could lead to a radiological emergency, has occurred).
NUREG-0654 states that the Initiating Conditions form the basis for establishment by a licensee of the specific plant instrumentation readings (as applicable) which, if exceeded, would initiate the emergency classification.
Thus, it is the specific instrument readings that would be the EALs. Considerations for the assignment of a particular Initiating Condition to an emergency classification level are discussed in Section 3. 2.3 EMERGENCY ACTION LEVEL (EAL) A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level. Discussion:
EAL statements may utilize a variety of criteria including instrument readings and status indications; observable events; results of calculations and analyses; entry into particular procedures; and the occurrence of natural phenomena.
2.4 FISSION PRODUCT BARRIER THRESHOLD A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Discussion:
Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment.
This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.
The primary fission product barriers are:
Fuel Clad
Reactor Coolant System (RCS)
Containment Upon determination that one or more fission product barrier thresholds have been exceeded, the combination of barrier loss and/or potential loss thresholds is compared to the fission product barrier IC/EAL criteria to determine the appropriate ECL. In some accident sequences, the ICs and EALs presented in the Abnormal Radiation 6
Levels/ Radiological Effluent (R) Recognition Category will be exceeded at the same time, or shortly after, the loss of one or more fission product barriers.
This redundancy is intentional as the former I Cs address radioactivity releases that result in certain offsite doses from whatever cause, including events that might not be fully encompassed by fission product barriers (e.g., spent fuel pool accidents, design containment leakage following a LOCA, etc.). 7 _J 3 DESIGN OF THE NEI 99-01 EMERGENCY CLASSIFICATION SCHEME ADOPTED BY MNGP 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS {ECLs) An effective emergency classification scheme must incorporate a realistic and accurate assessment of risk, both to plant workers and the public. There are obvious health and safety risks in underestimating the potential or actual threat from an event or condition; however, there are also risks in overestimating the threat as well (e.g., harm that may occur during an evacuation).
The NEI 99-0lemergency classification scheme attempts to strike an appropriate balance between reasonably anticipated event or condition consequences, potential accident trajectories, and risk avoidance or minimization.
NSPM has adopted the NEI 99-01 scheme, adding site-specific information as appropriate.
This section discusses the background for development of the NEI 99-01 scheme and adds MNGP specific details where appropriate.
There are a range of "non-emergency events" reported to the US Nuclear Regulatory Commission (NRC) staff in accordance with the requirements of 10 CFR § 50. 72. Guidance concerning these reporting requirements, and example events, are provided in NUREG-1022.
Certain events reportable under the provisions of 10 CFR § 50.72 may also require the declaration of an emergency.
In order to align each Initiating Conditions (IC) with the appropriate ECL, it was necessary to determine the attributes of each ECL. The goal of this process is to answer the question, "What events or conditions should be placed under each ECL ?" The following sources provided information and context for the development ofECL attributes.
Assessments of the effects and consequences of different types of events and conditions
Typical abnormal and emergency operating procedure setpoints and transition criteria
Typical Technical Specification limits and controls
Radiological Effluent Technical Specifications (RETS)/Offsite Dose Calculation Manual (ODCM) radiological release limits
Review of selected Updated Final Safety Analysis Report (UFSAR) accident analyses
Environmental Protection Agency (EPA) Protective Action Guidelines (PAGs)
NUREG 0654, Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants
Industry Operating Experience
Input from industry subject matter experts and MNGP staff The following ECL attributes were created to aid in the development of I Cs and Emergency Action Levels (EALs). These attributes may be useful in briefing and training settings (e.g., helping an Emergency Director understand why a particular condition is classified as an Alert). The attributes of each ECL are presented below. 8 3.1.1 Notification of Unusual Event (NUE) A Notification of Unusual Event, as defined in section 2.1.1, includes but is not limited to an event or condition that involves: (A)A precursor to a more significant event or condition. (B) A minor loss of control of radioactive materials or the ability to control radiation levels within the plant. (C) A consequence otherwise significant enough to warrant notification to local, State and Federal authorities.
3.1.2 Alert An Alert, as defined in section 2.1.2, includes but is not limited to an event or condition that involves: (A)A loss or potential loss of either the fuel clad or Reactor Coolant System (RCS) fission product barrier. (B) An event or condition that significantly reduces the margin to a loss or potential loss of the fuel clad or RCS fission product barrier. (C) A significant loss of control of radioactive materials resulting in an inability to control radiation levels within the plant, or a release of radioactive materials to the environment that could result in doses greater than 1 % of an EPA PAG at or beyond the site boundary. (D)A HOSTILE ACTION occurring within the OWNER CONTROLLED AREA, including those directed at an Independent Spent Fuel Storage Installation (ISFSI). 3.1.3 Site Area Emergency (SAE) A Site Area Emergency, as defined in section 2.1.3, includes but is not limited to an event or condition that involves: (A)A loss or potential loss of any two fission product barriers -fuel clad, RCS and/or containment. (B) A precursor event or condition that may lead to the loss or potential loss of multiple fission product barriers within a relatively short period of time. Precursor events and conditions of this type include those that challenge the monitoring and/or control of multiple safety systems. ( C) A release of radioactive materials to the environment that could result in doses greater than 10% of an EPA PAG at or beyond the site boundary. (D)A HOSTILE ACTION occurring within the plant PROTECTED AREA. 9 3.1.4 General Emergency (GE) A General Emergency, as defined in section 2.1.4, includes but is not limited to an event or condition that involves: (A) Loss of any two fission product barriers AND loss or potential loss of the third barrier -fuel clad, RCS and/or containment. (B) A precursor event or condition that, unmitigated, may lead to a loss of all three fission product barriers.
Precursor events and conditions of this type include those that lead directly to core damage and loss of containment integrity. (C) A release ofradioactive materials to the environment that could result in doses greater than an EPA PAG at or beyond the site boundary. (D)A HOSTILE ACTION resulting in the loss of key safety functions (reactivity control, core cooling/RPV water level or RCS heat removal) or damage to spent fuel. 3.1.5 Risk-Informed Insights Emergency preparedness is a defense-in-depth measure that is independent of the assessed risk from any particular accident sequence; however, the development of an effective emergency classification scheme can benefit from a review of risk-based assessment results. To that end, the development and assignment of certain ICs and EALs also considered insights from several site-specific probabilistic safety assessments (PSA -also known as probabilistic risk assessment, PRA). Some generic insights from this review included:
1. Accident sequences involving a prolonged loss of all AC power are significant contributors to core damage frequency.
For this reason, a loss of all AC power for greater than 15 minutes, with the plant at or above Hot Shutdown, was assigned an ECL of Site Area Emergency.
Precursor events to a loss of all AC power were also included as an Unusual Event and an Alert. A station blackout coping analyses performed in response to 10 CFR § 50.63 and Regulatory Guide 1.155, Station Blackout, was used to determine a time-based criterion to demarcate between a Site Area Emergency and a General Emergency.
The time dimension is critical to a properly anticipatory emergency declaration since the goal is to maximize the time available for State and local officials to develop and implement offsite protective actions. For MNGP, the coping analysis determined that MNGP is a four (4) hour coping plant. This provides the basis for the time-based demarcation criterion between a Site Area Emergency and a General Emergency for MNGP 2. For severe core damage events, uncertainties exist in phenomena important to accident progressions leading to containment failure. Because of these uncertainties, predicting the status of containment integrity may be difficult under severe accident conditions.
This is why maintaining containment integrity alone following sequences leading to severe core damage is an insufficient basis for not escalating to a General Emergency.
10
3. PSAs indicated that leading contributors to latent fatalities were sequences involving a containment bypass, a large Loss of Coolant Accident (LOCA) with early containment failure, a Station Blackout lasting longer than the site-specific coping period, and a reactor coolant pump seal failure. The generic EAL methodology needs to be sufficiently rigorous to address these sequences in a timely fashion. 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS The NEI 99-01 methodology adopted by MNGP makes use of symptom-based, based and event-based ICs and EALs. Each type is discussed below. Symptom-based ICs and EALs are parameters or conditions that are measurable over some range using plant instrumentation (e.g., core temperature, reactor coolant level, radiological effluent, etc.). When one or more of these parameters or conditions are normal, reactor operators will implement procedures to identify the probable cause(s) and take corrective action. Fission product barrier-based ICs and EALs are the subset of symptom-based EALs that refer specifically to the level of challenge to the principal barriers against the release of radioactive material from the reactor core to the environment.
These barriers are the fuel cladding, the reactor coolant system pressure boundary, and the containment.
The based ICs and EALs consider the level of challenge to each individual barrier -potentially lost and lost -and the total number of barriers under challenge.
Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance.
These include the failure of an automatic reactor scram/trip to shut down the reactor, natural phenomena (e.g., an earthquake), or man-made hazards such as a toxic gas release. 3.3 MNGP SPECIFIC DESIGN CONSIDERATIONS MNGP uses a single cycle, forced circulation, low power density boiling water reactor (BWR). General Electric Company designed the plant and supplied the nuclear steam supply system (NSSS), the initial reactor fuel, and turbine-generator unit and its related systems. This design is identified as "BWR-3" by General Electric.
ICs and EAL thresholds for a BWR NSSS have been appropriately incorporated into the MNGP emergency classification scheme. The reactor coolant system (also called the reactor primary system) includes the reactor vessel; the 2-loop reactor coolant recirculation system with its pumps, pipes and valves; the main steam piping up to the main steam isolation valves; safety/relief valves; and the reactor auxiliary systems piping. The reactor vessel contains the reactor core and supporting structure, steam separator and dryer assemblies, jet pumps, control rod guide tubes, and the Reactor Feedwater, Emergency Core Cooling System (ECCS), and Standby Liquid Control System spargers.
The Primary Containment System, consisting of a steel light-bulb-shaped drywell, a steel doughnut-shaped pressure suppression chamber, and interconnecting vent pipes, provides the first containment barrier surrounding the reactor vessel and reactor primary system. The primary containment system is designed to accommodate the pressures, 11 temperatures, and hydrodynamic loads which would result from, or occur subsequent to a postulated loss-of-coolant accident (LOCA) within the primary containment and safety relief valve operations.
Any leakage from the Primary Containment System is to the Secondary Containment System, consisting of the reactor building, the plant Standby Gas Treatment System, and the plant main stack. The primary safeguards functions of the secondary containment are to minimize ground level release of airborne radioactive materials, and to provide for controlled, filtered, elevated release of secondary containment atmosphere under postulated design basis accident conditions.
3.4 ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION The scheme's generic information is organized by Recognition Category in the following order.
R -Abnormal Radiation Levels I Radiological Effluent-Section 6
C -Cold Shutdown I Refueling System Malfunction
-Section 7
E -Independent Spent Fuel Storage Installation (ISFSI) -Section 8
F -Fission Product Barrier -Section 9
H -Hazards and Other Conditions Affecting Plant Safety -Section 10
S -System Malfunction
-Section 11 Each Recognition Category section contains a matrix showing the ICs and their associated emergency classification levels. The following information and guidance is provided for each IC:
ECL -the assigned emergency classification level for the IC.
Initiating Condition
-provides a summary description of the emergency event or condition.
Operating Mode Applicability
-Lists the modes during which the IC and associated EAL(s) are applicable (i.e., are to be used to classify events or conditions).
Emergency Action Level(s)-Provides reports and indications that are considered to meet the intent of the IC. For Recognition Category F, the fission product barrier thresholds are presented in a table applicable to MNGP, and arranged by fission product barrier and the degree of barrier challenge (i.e., potential loss or loss). This presentation method shows the synergism among the thi:esholds, and supports accurate assessments.
Basis -Provides background information that explains the intent and application of the IC and EALs. In some cases, the basis also includes relevant source information and references.
3.5 IC AND EAL MODE APPLICABILITY The NEI 99-01 emergency classification scheme adopted by MNGP was developed 12 recognizing that the applicability of ICs and EALs will vary with plant mode. For example, some symptom-based ICs and EALs can be assessed only during the power operations, startup, or hot shutdown modes of operation when all fission product barriers are in place, and plant instrumentation and safety systems are fully operational.
In the cold shutdown and refueling modes, different symptom-based I Cs and EALs will come into play to reflect the opening of systems for routine maintenance, the unavailability of some safety system components and the use of alternate instrumentation.
The following table shows which Recognition Categories are applicable in each plant mode. The ICs and EALs for a given Recognition Category are applicable in the indicated modes. MODE APPLICABILITY MATRIX FOR MNGP Category Mode R c E F H s Power Operations x x x x x Startup x x x x x Hot Shutdown x x x x x Cold Shutdown x x x x Refueling x x x x Defueled x x x x MNGP Operating Modes Reactor Mode Average Reactor Mode Title Switch Position Coolant Temperature
(°F) 1 Power Operation Run NA 2 Startup Refuel(a) or Startup/Hot Standby NA 3 Hot ShutdownCa)
Shutdown >212 4 Cold ShutdownCa)
Shutdown ::::;212 5 RefuelingCb)
Shutdown or Refuel NA Defueled (None): All fuel removed from the reactor vessel (i.e., full core offload during refueling or extended outage). (a) All reactor vessel head closure bolts fully tensioned. (b) One or more reactor vessel head closure bolts less than fully tensioned.
13 4 MNGP SCHEME DEVELOPMENT GUIDANCE 4.1 GENERAL IMPLEMENTATION GUIDANCE MNGP ICs and EALs were developed to be unambiguous and readily assessable.
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 MNGP ICs and EALs were developed with input from key stakeholders such as Operations, Training, Radiation Protection, Chemistry, and Engineering.
MNGP specific indications, parameters, and values are consistent with licensing basis documents, plant procedures, training, calculations, and drawings.
Useful acronyms and abbreviations associated with the MNGP emergency classification scheme are presented in Appendix A, Acronyms and Abbreviations.
Those specific to MNGP were included to be consistent with site terminology, site procedures, and training.
Many words or terms used in the MNGP emergency classification scheme have specific definitions.
These words and terms are identified by being set in all capital letters (i.e., ALL CAPS). The definitions are presented in Appendix B, Definitions.
4.2 CRITICAL CHARACTERISTICS When crafting the scheme, MNGP ensured that certain critical characteristics were met. These critical characteristics are listed below.
The ICs, EALs, Operating Mode Applicability criteria, Notes and Basis information are consistent with industry guidance; while the actual wording may be different from NEI 99-01 Revision 6, the classification intent is maintained.
With respect to Recognition Category F, the MNGP scheme includes a user-aid to facilitate timely and accurate classification of fission product barrier losses and/or potential losses. The user-aid logic is consistent with the classification logic presented in Section 9.
The ICs, EALs, Operating Mode Applicability criteria, Notes and Basis information are technically complete and accurate (i.e., they contain the information necessary to make a correct classification).
EAL statements use objective criteria and observable values.
ICs, EALs, Operating Mode Applicability and Note statements and formatting consider human factors and are user-friendly.
The scheme facilitates upgrading of the emergency classification where necessary.
The scheme facilitates classification of multiple concurrent events or conditions.
14 r----------4.3 INSTRUMENTATION USED FOR EALS MNGP EAL thresholds utilize instrumentation that is reliable and routinely maintained in accordance with site programs and procedures.
Alarms referenced in EAL statements are those that are the most operationally significant for the described event or condition.
MNGP personnel have ensured that specified values used as EAL setpoints are within the calibrated range of the referenced instrumentation, and consider any automatic instrumentation functions that may impact accurate EAL assessment.
In addition, EAL setpoint values do not use terms such as "off-scale low" or "off-scale high" since that type of reading may not be readily differentiated from an instrument failure. If instrumentation failures occur that have EALs associated with them (e.g., process radiation monitors) compensatory means of implementation may be used as described in plant procedures.
4.4 PRESENTATION OF SCHEME INFORMATION TO USERS The US Nuclear Regulatory Commission (NRC) expects licensees to establish and maintain the capability to assess, classify and declare an emergency condition promptly within 15 minutes after the availability of indications to plant operators that an emergency action level has been, or may be, exceeded.
The MNGP emergency classification procedure and user aid (EAL Matrix) have been developed to facilitate accurate and timely classification.
To this end, the following points have been considered.
The first users of an emergency classification procedure are the operators in the Control Room. During the allowable classification time period, they may have responsibility to perform other critical tasks, and will likely have minimal assistance in making a classification assessment.
As an emergency situation evolves, members of the Control Room staff are likely to be the first personnel to notice a change in plant conditions.
They can assess the changed conditions and, when warranted, recommend a different emergency classification level to the Technical Support Center (TSC) and/or Emergency Operations Facility (EOF).
Emergency Directors in the TSC and/or EOF will have more opportunity to focus on making an emergency classification, and will probably have advisors from Operations available to help them. The MNGP emergency classification scheme information for end users is presented in a manner with which licensed operators are most comfortable.
Input from the Operations and Operations Training Departments has been used to assist in the development of readily usable and easily understood classification tools (e.g., a procedure and EAL Matrix). The MNGP EAL Matrix contains all the information necessary to make a correct emergency classification.
MNGP EAL Matrix information includes the I Cs, Operating Mode Applicability criteria, EALs and Notes. Notes are adequately captured on the EAL Matrix and pointed to by each applicable EAL. Basis information is not included on the MNGP EAL Matrix but it is readily available to emergency classification decision-15 makers. MNGP has developed two matrices -one for use during power operations, startup and hot conditions, and another for cold shutdown and refueling conditions.
4.5 INTEGRATION OF ICs/EALs WITH PLANT PROCEDURES A rigorous integration ofIC and EAL references into plant operating procedures is not recommended.
This approach would greatly increase the administrative controls and workload for maintaining procedures.
On the other hand, performance challenges may occur ifrecognition of meeting an IC or EAL is based solely on the memory of a licensed operator or an Emergency Director, especially during periods of high stress. Visual cues (e.g., a step, note, caution, etc.) are included in plant procedures (including emergency operating procedures, abnormal operating procedures, alarm response procedures, and normal operating procedures), as appropriate, alerting the reader/user to consult the site emergency classification procedure.
4.6 BASIS DOCUMENT A basis document is an integral part of an emergency classification scheme. The material in this document supp01ts proper emergency classification decision-making by providing informing background and development information in a readily accessible format. It can be referred to in training situations and when making an actual emergency classification, if necessary.
The document is also useful for establishing configuration management controls for BP-related equipment and explaining an emergency classification to offsite authorities.
The content of the MNGP.basis document includes, at a minimum, the following:
An MNGP Mode Applicability Matrix and description of operating modes (Section 3.5).
A discussion of the emergency classification and declaration process (Section 5).
Each Initiating Condition along with the associated EALs or fission product barrier thresholds, Operating Mode Applicability, Notes and Basis information.
A listing of acronyms and defined terms, similar to that presented in Appendices A and B, respectively.
This material may be edited as needed to align with site-specific characteristics.
The MNGP Basis section does not contain information that could modify the meaning or intent of the associated IC or EAL. Information in the Basis is used only to clarify and inform decision-making for an emergency classification.
Basis information is readily available to be referenced, if necessary, by the Emergency Director.
A copy of the MNGP basis document is maintained in the appropriate emergency response facilities.
Because the information in a basis document can affect emergency classification decision-making (e.g., the Emergency Director refers to it during an event), changes to the MNGP basis document will be evaluated in accordance with the provisions of 10 CPR 50.54(q).
16 4.7 EAL/THRESHOLD REFERENCES TO MNGP AOP AND EOP SETPOINTS/CRITERIA The criteria/values used in several EALs and fission product barrier thresholds were derived from MNGP's AOPs and EOPs. This approach is intended to maintain good alignment between operational diagnoses and emergency classification assessments.
MNGP has verified that appropriate administrative controls are in place to ensure that a subsequent change to an AOP or EOP is screened to determine if an evaluation pursuant to 10 CFR 50.54(q) is required.
17 5 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 5.1 GENERAL CONSIDERATIONS When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basis information.
In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL. NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, Interim Staff Guidance, Emergency Planning for Nuclear Power Plants. All emergency classification assessments should be based upon valid indications, reports or conditions.
A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy.
For example, validation could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.
The validation of indications should be completed in a manner that supports timely emergency declaration.
For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.
A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that 1) the activity proceeds as planned and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component.
In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected.
Events or conditions of this type may be subject to the reporting requirements of 10 § CFR 50.72. The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.); the EAL and/or the associated basis discussion will identify the necessary analysis.
In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be 18 exceeded (i.e., this is the time that the EAL information is first available).
The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).
While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary.
The MNGP scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition.
A similar provision is incorporated into the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier. 5.2 CLASSIFICATION METHODOLOGY To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded.
The evaluation of an EAL(s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, then the IC is considered met and the associated ECL is declared in accordance with plant procedures.
When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock." For a full discussion of this timing requirement, refer to ISG-01. 5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared.
For example:
If an Alert EAL and a Site Area Emergency EAL are met 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 an Alert should be declared.
Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification ofNRC Guidance for Emergency Notifications During Quickly Changing Events. 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable.
If 19 an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared).
Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.
For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response.
In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 5.5 CLASSIFICATION OF IMMINENT CONDITIONS Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT).
If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.
5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING Once a classification level is declared, no downgrade to a lower classification will be allowed. The MNGP Emergency Plan and classification EPIPs provide the applicable guidance for transition to Termination and/or Recovery.
Guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02. 5.7 CLASSIFICATION OF SHORT-LIVED EVENTS As discussed in Section 3.2, event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance.
By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed.
If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration.
Examples of such events include an earthquake or a failure of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip.
5.8 CLASSIFICATION OF TRANSIENT CONDITIONS Many of the I Cs and/or EALs contained in this document employ time-based criteria.
These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted.
In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to 20 be met for a brief period of time (e.g., a few seconds to a few minutes).
The following guidance should be applied to the classification of these conditions.
EAL momentarily met during expected plant response -In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures.
EAL momentarily met but the condition is corrected prior to an emergency declaration
-If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required.
For illustrative purposes, consider the following example. An A TWS occurs and the EOPs direct lowering RPV water level to the top of active fuel or below in order to reduce reactor power. RPV water level is then controlled between the high level trip set point and the Minimum Steam Cooling RPV Water Level (MSCRWL), which challenges core cooling and the Fuel Clad barrier. In this situation, the immediate need to take manual action to reduce reactor power is the higher priority and the classification should be based on the ATWS only. It is important to stress that the 15-minute emergency classification assessment period is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event; emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations where an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration.
This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.
5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition.
This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery.
This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process. In these cases, no emergency declaration is waiTanted; however, the guidance contained in NUREG-1022 is applicable.
Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 within one hour of the discovery of the undeclared event or condition.
The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.
21 5.10 RETRACTION OF AN EMERGENCY DECLARATION Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022.
22 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS GENERAL EMERGENCY RGl Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE. Op. Modes: All RG2 Spent fuel pool level cannot be restored to at least 15 .25' for 60 minutes or longer. Op. Modes: All SITE AREA EMERGENCY RSl Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE. Op. Modes: All RS2 Spent fuel pool level at 15.25'. Op. Modes: All 23 ALERT RAl Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. Op. Modes: All RA2 Significant lowering of water level above, or damage to, irradiated fuel. Op. Modes: All RA3 Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Op. Modes: All UNUSUAL EVENT RUl Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer. Op. Modes: All RU2 UNPLANNED loss of water level above irradiated fuel. Op. Modes: All RG1 ECL: General Emergency Initiating Condition:
Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE. Operating Mode Applicability:
All Emergency Action Levels: (RG 1.1 or RG 1.2 or RG 1.3) Notes:
The Emergency Director should declare the General Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
The pre-calculated effluent monitor values presented in EAL RG 1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
RG 1.1 Reading on the following radiation monitor greater than the reading shown for 15 minutes or longer: I Stack Effluent Monitor (Ch A or B) I 8 E+07 µCi/sec RG 1.2 Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY.
RGl.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Basis:
Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.
Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.
SITE BOUNDARY:
For Dose Assessment and Protective Action Recommendation purposes the SITE BOUNDARY is the closest distance at which members of the public would be exposed to a radioactive release. The SITE BOUNDARY for liquid releases ofradioactive material is defined in ODCM-02.01 (LIQUID EFFLUENTS).
The SITE BOUNDARY for gaseous releases of radioactive material is defined in ODCM-03.01 (GASEOUS EFFLUENTS).
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.
Releases of this magnitude will require implementation of protective actions for the public. 24 Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
MNGP Basis Reference(s):
1. Monticello Calculation CA 04-199, Methodology Used to Derive Radiation Monitor Readings for NEI 99-01 Rev 6 2. ODCM-01.01 (OFF-SITE DOSE CALCULATION MANUAL (ODCM) INTRODUCTION)
3. ODCM-03.01 (GASEOUS EFFLUENTS), Section 2.7.l, Gaseous Effluents Bases 4. ODCM-03.01 (GASEOUS EFFLUENTS), Figure 1, Monticello Nuclear Generating Plant Site Boundary for Gaseous Effluents
5. A.2-807 (OFF-SITE DOSE ASSESSMENT AND PROTECTIVE ACTION RECOMMENDATIONS)
6. Ops Man B.05.11-03 (PROCESS RADIATION MONITORING INSTRUMENTATION AND CONTROLS) 25 RG2 ECL: General Emergency Initiating Condition:
Spent fuel pool level cannot be restored to at least 15.25' for 60 minutes or longer. Operating Mode Applicability:
All Emergency Action Levels: Note: The Emergency Director should declare the General Emergency promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
RG2.l Spent fuel pool level cannot be restored to at least 15.25' for 60 minutes or longer. Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.
It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.
MNGP Basis Reference(s):
1. Ops Man B.02.01-03 (FUEL POOL COOLING-INSTRUMENTATION AND CONTROLS)
2. Ops Man B.02.01-05 (FUEL POOL COOLING SYSTEM OPERATION) 26 RS1 ECL: Site Area Emergency Initiating Condition:
Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE. Operating Mode Applicability:
All Emergency Action Levels: (RSl.l or RSl.2 or RSl.3) Notes:
The Emergency Director should declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
The pre-calculated effluent monitor values presented in EAL RS 1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
RS 1.1 Reading on the following radiation monitor greater than the reading shown for 15 minutes or longer: I Stack Effluent Monitor (Ch A or B) I 8 E+06 µCi/sec RS 1.2 Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY.
RS 1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Basis:
Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes . or longer.
Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.
SITE BOUNDARY:
For Dose Assessment and Protective Action Recommendation purposes the SITE BOUNDARY is the closest distance at which members of the public would be exposed to a radioactive release. The SITE BOUNDARY for liquid releases ofradioactive material is defined in ODCM-02.01 (LIQUID EFFLUENTS).
The SITE BOUNDARY for gaseous releases of radioactive material is defined in ODCM-03.01 (GASEOUS EFFLUENTS).
This IC addresses a release of gaseous radioactivity that results iii projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PA Gs). It includes both monitored and un-monitored releases.
Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. 27 Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RGI. MNGP Basis Reference(s):
1. Monticello Calculation CA 04-199, Methodology Used to Derive Radiation Monitor Readings for NEI 99-01 Rev 6 2. ODCM-01.01 (OFF-SITE DOSE CALCULATION MANUAL (ODCM) INTRODUCTION)
3. ODCM-03.01 (GASEOUS EFFLUENTS), Section 2.7.1, Gaseous Effluents Bases 4. ODCM-03.01 (GASEOUS EFFLUENTS), Figure 1, Monticello Nuclear Generating Plant Site Boundary for Gaseous Effluents
5. A.2-807 (OFF-SITE DOSE ASSESSMENT AND PROTECTIVE ACTION RECOMMENDATIONS)
6. Ops Man B.05.11-03 (PROCESS RADIATION MONITORING INSTRUMENTATION AND CONTROLS) 28 RS2 ECL: Site Area Emergency Initiating Condition:
Spent fuel pool level at 15.25'. Operating Mode Applicability:
All Emergency Action Levels: RS2.1 Lowering of spent fuel pool level to 15.25'. Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.
Escalation of the emergency classification level would be via IC RG 1 or RG2. MNGP Basis Reference(s):
1. Ops Man B.02.01-03 (FUEL POOL COOLING -INSTRUMENTATION AND CONTROLS)
2. Ops Man B.02.01-05 (FUEL POOL COOLING SYSTEM OPERATIONS) 29
,-------RA1 ECL: Alert Initiating Condition:
Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. Operating Mode Applicability:
All Emergency Action Levels: (RAl.1 or RAl.2 or RAl.3 or RAl.4) Notes:
The Emergency Director should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
The pre-calculated effluent monitor values presented in EAL RAl .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
RAl.1 Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: Gaseous Effluent Monitors Stack Effluent Monitor (Ch A or B) 8 E+05 µCi/sec RB Vent Effluent Monitor (Ch A or B) 6 E+05 µCi/sec Liquid Effluent Monitors Discharge Canal 2000 cps Service Water 700 cps TBNWS 9 E+04 cpm RAl .2 Dose assessment using actual meteorology indicates greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY.
RAl .3 Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for one hour of exposure.
RAl.4 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.
Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.
30 Basis: SITE BOUNDARY:
For Dose Assessment and Protective Action Recommendation purposes the SITE BOUNDARY is the closest distance at which members of the public would be exposed to a radioactive release. The SITE BOUNDARY for liquid releases of radioactive material is defined in ODCM-02.01 (LIQUID EFFLUENTS).
The SITE BOUNDARY for gaseous releases of radioactive material is defined in ODCM-03.01 (GASEOUS EFFLUENTS).
This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PA Gs). It includes both monitored and un-monitored releases.
Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1 % of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC RS 1. MNGP Basis Reference(s):
1. Monticello Calculation CA 04-199, Methodology Used to Derive Radiation Monitor Readings for NEI 99-01 Rev 6 2. ODCM-01.01 (OFF-SITE DOSE CALCULATION MANUAL (ODCM) INTRODUCTION)
3. ODCM-02.01 (LIQUID EFFLUENTS), Figure 1, Monticello Nuclear Generating Plant Site Boundary for Liquid Effluents
4. ODCM-03.01 (GASEOUS EFFLUENTS), Section 2.7.1, Gaseous Effluents Bases 5. ODCM-03.01 (GASEOUS EFFLUENTS), Figure 1, Monticello Nuclear Generating Plant Site Boundary for Gaseous Effluents
6. A.2-807 (OFF-SITE DOSE ASSESSMENT AND PROTECTIVE ACTION RECOMMENDATIONS)
7. Ops Man B.05.11-03 (PROCESS RADIATION MONITORING INSTRUMENTATION AND CONTROLS) 31 RA2 ECL: Alert Initiating Condition:
Significant lowering of water level above, or damage to, in-adiated fuel. Operating Mode Applicability:
All Emergency Action Levels: (RA2.l or RA2.2 or RA2.3) RA2.1 Uncovery of irradiated fuel in the REFUELING PATHWAY. RA2.2 Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of the following radiation monitors:
A-1 1027 RB NE Low A-2 1027 RB N High A-3 1027 RB W Stairway Monitor RM-17-452A Reactor Building Ventilation Exhaust Plenum Monitor Ch A RM-17-452B Reactor Building Ventilation Exhaust Plenum Monitor Ch B RM-17-453A Fuel Pool Radiation Monitor Ch A RM-17-453B Fuel Pool Radiation Monitor Ch B RA2.3 Lowering of spent fuel pool level to 24.75'. Basis: Alarm/Trip 20 mR/hr 200 mR/hr 100 mRlhr 26 mR/hr 26 mR/hr 50 mR/hr 50 mR/hr REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool, or fuel transfer canal. This IC addresses events that have caused IMMINENT or actual damage to an in-adiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment.
As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC EUl. Escalation of the emergency would be based on either Recognition Category R or C ICs. EALRA2.1 This EAL escalates from RU2 in that the loss oflevel, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of in-adiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters.
Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.
32 While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered.
To the degree possible, readings should be considered in combination with other available indications of inventory loss. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. EALRA2.2 This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly.
A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).
EALRA2.3 Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via ICs RSl or RS2 MNGP Basis Reference(s):
1. Ops Man B.05.12 (AREA RADIATION MONITORING SYSTEM) 2. Ops Man B.05.11 (PROCESS RADIATION MONITORING SYSTEM) 3. USAR Section 7.5.2, Plant Instrumentation and Control Systems -Process Radiation Monitoring System 4. USAR Section 7.5.3, Plant Instrumentation and Control Systems-Area Radiation Monitoring System 5. C.6-004-A-01 (REFUELING FLOOR AREA HI RADIATION)
6. C.6-005-A-Ol (REAC BLDG VENT & F P RAD CH A-HI/LO) 7. C.6-005-A-02 (REAC BLDG VENT & F P RAD CH B-HI/LO) 8. Ops Man B.02.01-03 (FUEL POOL COOLING-INSTRUMENTATION AND CONTROLS)
9. Ops Man B.02.01-05 (FUEL POOL COOLING SYSTEM OPERATIONS)
10. 1024 (AREA RADIATION MONITOR CALIBRATION)
11. NX-9321-86, Spent Fuel Pool 33
12. CAP 1543790, No Technical Basis for Area Monitor Alarm Setpoints 34 RA3 ECL: Alert Initiating Condition:
Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability:
All Emergency Action Levels: (RA3.l or RA3.2) Note: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.
RA3.l Dose rate greater than 15 mR/hr in ANY of the following areas:
Control Room (A-20 Control Room Low Range)
Central Alarm Station (by survey)
Secondary Alarm Station (by survey) RA3.2 An UNPLANNED event results in radiation levels that prohibit or impede access to any of the Table Hl plant rooms or areas: Table Hl Building Rooms Applicable Mode(s) Reactor Building All All Turbine Building All All Intake Structure All All Basis: UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown.
As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable. For EAL RA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits). An emergency declaration is not warranted if any of the following conditions apply. 35
The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).
The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action. Escalation of the emergency classification level would be via Recognition Category R, C or F I Cs. MNGP Basis Reference(s):
1. GDC 19, Control Room 2. NUREG-0737, "Clarification of TMI Action Plan Requirements",Section III.D.3 3. Ops Man B.05.12 (AREA RADIATION MONITORING SYSTEM) 4. Ops Man C.l (STARTUP PROCEDURE)
5. Ops Man C.3 (SHUTDOWN PROCEDURE)
6. Ops Man C.5.1-1300 (SECONDARY CONTAINMENT CONTROL) 7. USAR Section .10.3.1.5.1, Plant Auxiliary Systems -Safe Shutdown Analysis, General 8. USAR Section 12.2, Plant Structures and Shielding-Plant Principal Structures and Foundations
9. ND-95208, Monticello Property Map 10. ND-95209, Monticello Main Plant Structures
11. 4AWI-08.04.01 (RADIATION PROTECTION PLAN) 12. A.2-401 (EMERGENCY EXPOSURE CONTROL) 36 RU1 ECL: Notification of Unusual Event Initiating Condition:
Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer. Operating Mode Applicability:
All Emergency Action Levels: (RUl.1 or RUl.2 or RUl.3) Notes:
The Emergency Director should declare the Unusual Event promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
RUl.l Reading on ANY of the following effluent radiation monitors greater than the listed values for 60 minutes or longer: Gaseous Effluent Monitors Stack Effluent Monitor (Ch A or B) 4 E+05 µCi/sec RB Vent Effluent Monitor (Ch A or B) 3 E+04 µCi/sec Liquid Effluent Monitors Discharge Canal 900 cps Service Water 300 cps TBNWS 4E+04 cpm RUl .2 Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer. RUl .3 Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODCM limits for 60 minutes or longer. Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release).
It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment.
Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases.
The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
37 Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Releases should not be prorated or averaged.
For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. EAL RUl .1 -This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.
EAL RUl .2 -This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas). EAL RUl.3 -This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.). Escalation of the emergency classification level would be via IC RAl. MNGP Basis Reference(s):
1. Monticello Calculation 04-199, Methodology Used to Derive Radiation Monitor Readings for NEI 99-01 Rev 6 2. ODCM-01.01 (OFF-SITE DOSE CALCULATION MANUAL (ODCM) INTRODUCTION)
3. ODCM-02.01 (LIQUID EFFLUENTS)
4. ODCM-03.01 (GASEOUS EFFLUENTS) 38 RU2 ECL: Notification of Unusual Event Initiating Condition:
UNPLANNED loss of water level above irradiated fuel. Operating Mode Applicability:
All Emergency Action Levels: RU2.1 a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:
Basis: AND Spent Fuel Pool low water level alarm Visual observation of an uncontrolled water level drop below a fuel pool skimmer surge tank inlet
Observation of water leakage into the drywell or the reactor building from piping penetrations surrounding the drywell. b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.
A-1 1027 RB NE Low A-2 1027 RB N High A-3 1027 RB W Stairway RM-17-453A or B Fuel Pool Monitoring System REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool, or fuel transfer canal. UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant. A water level decrease will be primarily determined by indications from available level instrumentation.
Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available).
A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.
The effects of planned evolutions should be considered.
For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly.
Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level. 39 I .. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level would be via IC RA2. MNGP Basis Reference(s):
1. Ops Man B.02.01-01 (FUEL POOL COOLING-FUNCTION
& GENERAL DESCRIPTION OF SYSTEM) 2. Ops Man B.05.12 (AREA RADIATION MONITORING SYSTEM) 3. Procedure OOOOJ, Operations Daily Log 4. Ops Man C.6-004-B-33 (FUEL POOL COOLING SYSTEM TROUBLE) 5. Ops Man C.6-065-A-06 (REFUELING HIGH LEAKAGE) 40 7 COLD SHUTDOWN/
REFUELING SYSTEM MALFUNCTION ICS/EALS GENERAL EMERGENCY CG 1 Loss of RPV inventory affecting fuel clad integrity with containment challenged.
Op. Modes: Cold Shutdown, Refueling SITE AREA EMERGENCY CSl Loss ofRPV inventory affecting core decay heat removal capability.
Op. Modes: Cold Shutdown, Refueling 41 ALERT CAl Loss of RPV inventory.
Op. Modes: Cold Shutdown, Refueling CA2 Loss of all offsite and all onsite AC power to essential buses for 15 minutes or longer. Op. Modes: Cold Shutdown, Refueling, Defueled CA3 Inability to maintain the plant in cold shutdown.
Op. Modes: Cold Shutdown, Refueling CA6 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Op. Modes: Cold Shutdown, Refueling UNUSUAL EVENT CUl UNPLANNED loss of RPV inventory for 15 minutes or longer. Op. Modes: Cold Shutdown, Refueling CU2 Loss of all but one AC power source to essential buses for 15 minutes or longer. Op. Modes: Cold Shutdown, Refueling, Defueled CU3 UNPLANNED rise in RCS temperature.
Op. Modes: Cold Shuidown, Refueling CU 4 Loss of Vital DC power for 15 minutes or longer. Op. Modes: Cold Shutdown, Refueling CU5 Loss of all onsite or offsite communications capabilities.
Op. Modes: Cold Shutdown, Refueling, Defueled CG1 ECL: General Emergency Initiating Condition:
Loss of RPV inventory affecting fuel clad integrity with containment challenged.
Operating Mode Applicability:
Cold Shutdown, Refueling Emergency Action Levels: (CGl.l or CGl.2) Note: The Emergency Director should declare the General Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.
CGl.l a. RPV level less than -126 in. (TAF) for 30 minutes or longer. AND b. ANY indication from the Containment Challenge Table Cl. CGl.2 a. RPV level cannot be monitored for 30 minutes or longer. AND b. Core uncovery is indicated by EITHER of the following:
AND
Refueling Floor radiation monitor reading greater than 3 R/hr
UNPLANNED rise in drywell floor or equipment drain sump levels of sufficient magnitude to indicate core uncovery c. ANY indication from the Containment Challenge Table Cl Containment Challenge Table Cl
SECONDARY CONTAINMENT not established*
Greater than or equal to 6% hydrogen and greater than or equal to 5% oxygen m primary containment
UNPLANNED rise in containment pressure of greater than 1.84 psig
Two or more Reactor Building areas exceed Max Safe Radiation Levels (C.5-1300, Table X) *If SECONDARY CONTAINMENT is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.
Basis: SECONDARY CONTAINMENT:
SECONDARY CONTAINMENT includes the Reactor Building (including the HPCI Building), the Standby Gas Treatment System, the Offgas Dilution Fans, and connecting pipes and ducts. SECONDARY CONTAINMENT is isolated along with an automatic initiation of the Standby Gas Treatment System to minimize radiological releases to the environment.
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may 42 be known or unknown. This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged.
This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.
With SECONDARY CONTAINMENT not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment.
If SECONDARY CONTAINMENT is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.
The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity.
It therefore represents a challenge to Containment integrity.
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment.
If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.
In EAL CG 1.2.b, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).
It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.
The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation.
If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.
MNGP Basis Reference(s):
43
1. Ops Man C.l (STARTUP PROCEDURE)
2. Ops Man C.4-B-04.01.F (LEAK INSIDE PRIMARY CONTAINMENT)
3. Ops Man C.6-004-B-13 (DRYWELL EQUIP DRAIN LEAK RATE HI) 4. Ops Man C.6-004-B-17 (DRYWELL FLOOR DRAIN SUMP HI) 5. Ops Man C.6-004-B-18 (DRYWELL EQUIP DRAIN LEAK RATE CHANGE HI) 6. Ops Man C.6-065-A-06 (REFUELING:
HIGH LEAKAGE) 7. Emergency Operating Procedure C.5.1-1100 (RPV CONTROL) 8. Emergency Operating Procedure C.5.1-1300 (SECONDARY CONTAINMENT CONTROL) 9. USAR Section 5 .2, Containment System -Primary Containment System 10. USAR Section 7.9-1, Plant Instrumentation and Control Systems-Accident Monitoring Instrumentation
11. MNGP Calculation CA-04-202, Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level 44 CS1 ECL: Site Area Eniergency Initiating Condition:
Loss of RPV inventory affecting core decay heat removal capability.
Operating Mode Applicability:
Cold Shutdown, Refueling Emergency Action Levels: (CSl.l or CSl.2 or CSl.3) Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.
CS 1.1 a. SECONDARY CONTAINMENT not established.
AND b. RPV level less than -4 7 in. CSl.2 a. SECONDARY CONTAINMENT established.
AND b. RPV level less than -126 in. (TAF). CSl.3 a. RPV level cannot be monitored for 30 minutes or longer. Basis: AND b. Core uncovery is indicated by EITHER of the following:
Refueling Floor radiation monitor reading greater than 3 R/hr
UNPLANNED rise in drywell floor or equipment drain sump levels of sufficient magnitude to indicate.core uncovery SECONDARY CONTAINMENT:
SECONDARY CONTAINMENT includes the Reactor Building (including the HPCI Building), the Standby Gas Treatment System, the Offgas Dilution Fans, and connecting pipes and ducts. SECONDARY CONTAINMENT is isolated along with an automatic initiation of the Standby Gas Treatment System to minimize radiological releases to the environment.
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses a significant and prolonged loss ofRPV inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. . Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.
45 Outage/shutdown contingency plans typically provide for re-establishing or verifying SECONDARY CONTAINMENT following a loss of heat removal or RCS inventory control functions.
The difference in the specified RCS/reactor vessel levels ofEALs CSl.1.b and CSl.2.b reflect the fact that with SECONDARY CONTAINMENT established, there is a lower probability of a fission product release to the environment.
In EAL CSl.3.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).
It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.
The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation.
If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative ofleakage from the RPV. These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.
Escalation of the emergency classification level would be via IC CGl or RGI. MNGP Basis Reference(s):
1. Ops Man B.01.01-03 (REACTOR AND VESSEL ASSEMBLY)
2. Ops Man B.05.06 (PLANT PROTECTION SYSTEM) 3. Ops Man C.l (STARTUP PROCEDURE)
4. Ops Man C.4-B.04.01.F (LEAK INSIDE PRIMARY CONTAINMENT)
5. Ops Man C.6-003-A-38 (REACTOR LOW LOW LEVEL) 6. Ops Man C.6-004-B-13 (DRYWELL EQUIP DRAIN LEVEL RATE HI) 7. Ops Man C.6-004-B-17 (DRYWELL FLOOR DRAIN SUMP HI LEVEL) 8. Ops Man C.6-004-B-18 (DRYWELL EQUIP DRAIN LEAK RATE CHANGE HI) 9. Tech Spec Table 3.3.5.1-1 (EMERGENCY CORE COOLING INSTRUMENTATION)
10. Tech Spec 3.4.4 (RCS OPERATIONAL LEAKAGE) 11. NX-7831-197-1, Reactor Vessel and Internals
12. MNGP Calculation CA-95-074, Low Low Reactor Water Level Group 1 and 3 Containment Isolation Setpoint Calculation 46
13. MNGP Calculation CA-04-202, Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level 47 CA1 ECL: Alert Initiating Condition:
Loss of RPV inventory.
Operating Mode Applicability:
Cold Shutdown, Refueling Emergency Action Levels: (CALI or CAl.2) Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
CAl.1 Loss ofRPV inventory as indicated by level less than -47 in. CAl.2 a. RPV level cannot be monitored for 15 minutes or longer Basis: AND b. UNPLANNED rise in drywell floor or equipment drain sump levels due to a loss ofRPV inventory.
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precu'rsor to a challenge to the fuel clad barrier).
This condition represents a potential substantial reduction in the level of plant safety. For EAL CAl .1, a lowering of water level below -47 in. (Low-Low ECCS actuation setpoint) indicates that operator actions have not been successful in restoring and maintaining RPV water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.
Although related, EAL CA 1.1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3. For EAL CAl.2, the inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation.
If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS 1 48 If the RPV inventory level continues to lower, then escalation to Site Area Emergency would be via IC CSL MNGP Basis Reference(s):
1. Ops Man B.O 1.01-03 (REACTOR AND VESSEL ASSEMBLY -INSTRUMENTATION AND CONTROLS)
2. Ops Man B.05.06 (PLANT PROTECTION SYSTEM) 3. Ops Man C.4-B.04.01.F (LEAK INSIDE PRIMARY CONTAINMENT)
4. Ops Man C.6-003-A-38 (REACTOR LOW LOW LEVEL) 5. Ops Man C.6-004-B-13 (DRYWELL EQUIP DRAIN LEAK RATE HI) 6. Ops Man C.6-004-B-17 (DRYWELL FLOOR DRAIN SUMP HI) 7. Ops Man C.6-004-B-18 (DRYWELL EQUIP DRAIN LEAK RATE CHANGE HI) 8. 9040 (TEMPORARY VESSEL LEVEL INSTRUMENTATION INSTALLATION AND RESTORATION)
9. Tech Spec Table 3.3.5.1-1 (EMERGENCY CORE COOLING SYSTEM INSTRUMENTATION)
10. Ops Man C.6-065-A-06 (REFUELING HIGH LEAKAGE) 49 CA2 ECL: Alert Initiating Condition:
Loss of all off site and all onsite AC power to essential buses for 15 minutes or longer. Operating Mode Applicability:
Cold Shutdown, Refueling, Defueled Emergency Action Levels: Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
CA2.1 Loss of ALL offsite and ALL onsite AC Power to essential buses 15 and 16 for 15 minutes or longer. Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. This EAL is indicated by the loss of all offsite and onsite AC power to the 4160V essential buses. Onsite sources include 11 and 12 Emergency Diesel Generators.
Offsite resources include 2R, IR, and lAR Transformers.
If power is available from these sources, but is not supplied to the 4160V buses for whatever reason, the condition is still considered a loss of power. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an essential bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC CS 1 or RS 1. MNGP Basis Reference(s):
1. USAR Section 8.2, Plant Electrical Systems -Transmission System 2. USAR Section 8.3, Plant Electrical Systems -Auxiliary Power System 3. USAR Section 8.4, Plant Electrical Systems -Plant Standby Diesel Generator System 4. USAR Figure 8.4-1, Diesel Generation System One Line Diagram 5. NF-36175, Single Line Diagram -Station Connection 50
6. Ops Man B.09.06-01 (4.16 KV STATION AUXILIARY -FUNCTION AND GENERAL DESCRIPTION OF SYSTEM) 7. Ops Man C.4-B.09.02.A (STATION BLACKOUT)
8. Ops Man C.4-B.09.02.B (LOSS OF NORMAL OFF-SITE POWER) 9. Ops Man C.4-B.09.06.C (LOSS OF BUS 15 OR BUS 16) 10. Tech Spec 3.8.2 (AC SOURCES-SHUTDOWN) 51 -____ _I CA3 ECL: Alert Initiating Condition:
Inability to maintain the plant in cold shutdown.
Operating Mode Applicability:
Cold Shutdown, Refueling Emergency Action Levels: (CA3.l or CA3.2) Note: The Emergency Director should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
CA3 .1 UNPLANNED rise in RCS temperature to greater than 212 °F for greater than the duration specified in Table C2. Table C2: RCS Heat-up Duration Thresholds RCS SECONDARY Heat-up Duration CONTAINMENT Not intact Not Established 0 minutes Established 20 minutes* Intact NIA 60 minutes*
If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
CA3.2 UNPLANNED RCS pressure rise greater than 10 psig. Basis: SECONDARY CONTAINMENT:
SECONDARY CONTAINMENT includes the Reactor Building (including the HPCI Building), the Standby Gas Treatment System, the Offgas Dilution Fans, and connecting pipes and ducts. SECONDARY CONTAINMENT is isolated along with an automatic initiation of the Standby Gas Treatment System to minimize radiological releases to the environment.
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when SECONDARY CONTAINMENT is established but the RCS is not intact. The 20-minute criterion was included to allow time for operator action to address the temperature increase.
52 The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of SECONDARY CONTAINMENT is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety. Finally, in the case where there is an increase in RCS temperature, the RCS is not intact and SECONDARY CONTAINMENT is not established, no heat-up duration is allowed (i.e., 0 minutes).
This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel. EAL CA3.2 provides a pressure-based indication of RCS heat-up. Escalation of the emergency classification level would be via IC CS 1 or RS 1. MNGP Basis Reference(s):
I. Ops Man B.01.01-03 (REACTOR AND VESSEL ASSEMBLY -INSTRUMENTATION AND CONTROLS)
2. Ops Man C.3 (SHUTDOWN PROCEDuRE)
3. MNGP Emergency Plan Table 13, Instruments Available For Monitoring Major Systems 4. USAR Section 7.4, Plant Instrumentation and Control Systems-Reactor Vessel Instrumentation
5. Tech Spec Table 1.1-1 (MODES) 6. Tech Spec 3.6.4.1 (SECONDARY CONTAINMENT)
7. Tech Spec 3.6.4.2 (SECONDARY CONTAINMENT ISOLATION VALVES (SCIVs)) 8. Tech Spec 3.6.4.3 (STANDBY GAS TREATMENT (SGT) SYSTEM) 53 CA6 ECL: Alert Initiating Condition:
Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability:
Cold Shutdown, Refueling Emergency Action Levels: CA6.l Basis: a. The occurrence of ANY of the following hazardous events:
Seismic event (Earthquake)
Internal or external flooding event
High winds or tornado strike
FIRE
EXPLOSION
River level greater than 919 ft el.
River level less than 900.5 ft el.
Other events with similar hazard characteristics as determined by the Shift Manager AND b. EITHER of the following:
Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.
The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode. EXPLOSION:
A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.
Such events may require a post-event inspection to determine if the attributes of an explosion are present. FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, 54 and therefore represents an actual or potential substantial degradation of the level of safety of the plant. The first threshold for EAL CA6. lb addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.
The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. The second threshold for EAL CA6.lb addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components.
Operators will make this determination based on the totality of available event and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC CS 1 or RS 1. MNGP Basis Reference(s):
1. Ops Man A.6 (ACTS OF NATURE) 2. Ops Man B.05.14 (SEISMIC MONITORING SYSTEM) 3. Ops Man B.05.16-01 (METEOROLOGICAL MONITORING-FUNCTION
& GENERAL DESCRJPTION OF SYSTEM) 4. Ops Man B.06.04 (CIRCULATING WATER SYSTEM) 5. C.4-B.05.14.A (EARTHQUAKE)
6. C.6-006-C-08 (EARTHQUAKE)
7. C.6-006-C-13 (OPERATIONAL BASIS EARTHQUAKE)
8. USAR Section 10.3, Plant Auxiliary Systems -Plant Service Systems 9. USAR Section 12.2, Plant Principal Structures and Foundations
10. USAR Appendix G, Chapter 3, Probable Maximum Flood Determination
11. USAR Table I.5-1, Location of High Energy Systems and Safe Shutdown Equipment by Volume 12. USAR Appendix J.4, Fire Protection Program -Safe Shutdown Analysis 13. ND-95208, Monticello Property Map 14. ND-95209, Monticello Main Plant Structures
15. 4 AWI-01.03.01 (QUALITY ASSURANCE PROGRAM BOUNDARY) 55

CU1 ECL: Notification of Unusual Event Initiating Condition:
UNPLANNED loss ofRPV inventory for I5 minutes or longer. Operating Mode Applicability:
Cold Shutdown, Refueling Emergency Action Levels: (CUI.I or CUI.2) Note: The Emergency Director should declare the Unusual Event promptly upon determining that I5 minutes has been exceeded, or will likely be exceeded.
CU I. I UNPLANNED loss of reactor coolant results in RPV level less than a procedurally required lower limit for I 5 minutes or longer. CUI.2 a. RPV level cannot be monitored.
AND b. UNPLANNED rise in drywell floor or equipment drain sump levels. Basis: UNPLANNED:
A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV 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 decrease RCS water inventory are carefully planned and controlled.
An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. EAL CUI. I recognizes that the minimum required RPV level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented.
This EAL is met ifthe minimum level, specified for the current plant conditions, cannot be maintained for I5 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.
The I5-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level. EAL CUI .2 addresses a condition where all means to determine RPV level have been lost. In, this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. 56 Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CAI or CA3. MNGP Basis Reference(s):
1. Ops Man B.01.01-06 (REACTOR AND VESSEL ASSEMBLY -FIGURES) 2. Ops Man C.3 (SHUTDOWN PROCEDURE)
3. 9001 (REACTOR WELL AND DRYER-SEPARATOR STORAGE POOL FILLING PROCEDURE)
4. 9006 (REACTOR WELL AND DRYER-SEPARATOR STORAGE POOL DRAINING PROCEDURE)
5. 9040 (TEMPORARY VESSEL LEVEL INSTRUMENTATION INSTALLATION AND RESTORATION)
6. C.5-1100 (RPV CONTROL) 7. Ops Man C.4-B.04.01.F (LEAK INSIDE PRIMARY CONTAINMENT)
8. Ops Man C.6-003-A-38 (REACTOR LOW LEVEL) 9. Ops Man C.6-004-B-13 (DRYWELL EQUIP DRAIN LEAK RATE HI) 10. Ops Man C.6-004-B-17 (DRYWELL FLOOR DRAIN SUMP HI LEVEL) 11. Ops Man C.6-004-B-18 (DRYWELL EQUIP DRAIN LEAK RATE CHANGE HI) 12. Ops Man C.6-065-A-06 (REFUELING HIGH LEAKAGE) 13. Technical Specification 3.9 (REFUELING OPERATIONS)
14. MNGP Calculation CA-95-074, Low Low Reactor Water Level Group 1 and 3 Containment Isolation Setpoint Calculation 57 CU2 ECL: Notification of Unusual Event Initiating Condition:
Loss of all but one AC power source to essential buses for 15 minutes or longer. Operating Mode Applicability:
Cold Shutdown, Refueling, Defueled Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
CU2.1 a. AC power capability to essential buses 15 and 16 is reduced to a single power source (Table Sl) for 15 minutes or longer. Basis: AND b. Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS. Table Sl IR Reserve Transformer lAR Reserve Transformer 2R Auxiliary Transformer
11 Emergency Diesel Generator
12 Emergency Diesel Generator SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
This IC describes a significant degradation of off site and onsite AC power sources (Table S 1) such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.
When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an essential bus. Some examples of this condition are presented below.
A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
A loss of emergency power sources (e.g., onsite diesel generators) with a single train of 58 essential buses being back-fed from an 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 in accordance with IC CA2. MNGP Basis Reference(s):
1. USAR Section 8.2, Plant Electrical Systems -Transmission System 2. USAR Section 8.3, Plant Electrical Systems -Auxiliary Power Systems 3. USAR Section 8.4, Plant Electrical Systems -Plant Standby Diesel Generator System 4. USAR Figure 8.4-1, Diesel Generation System One Line Diagram ' 5. NF-36175, Single Line Diagram-Station Connection
6. Ops Man B.09.06-01 (4.16 KV STATION AUXILIARY -FUNCTION AND GENERAL DESCRIPTION OF SYSTEM) 7. Ops Man C.4-B.09.02.A (STATION BLACKOUT)
8. Ops Man C.4-B.09.02.B (LOSS OF NORMAL OFF-SITE POWER) 9. Ops Man C.4-B.09.06.C (LOSS OF BUS 15 OR BUS 16) 10. Tech Spec 3.8.2 (AC SOURCES-SHUTDOWN) 59 CU3 ECL: Notification of Unusual Event Initiating Condition:
UNPLANNED rise in RCS temperature.
Operating Mode Applicability:
Cold Shutdown, Refueling Emergency Action Levels: (CU3.l or CU3.2) Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
CU3.l UNPLANNED rise in RCS temperature to greater than 212 °F. CU3.2 Loss of ALL RCS temperature and RPV level indication for 15 minutes or longer. Basis: UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and SECONDARY CONTAINMENT is not established during this event, the Emergency Director should also refer to IC CA3. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
EAL CU3.l involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications.
During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.
During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled.
A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.
EAL CU3.2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
60 Escalation to Alert would be via IC CAl based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
MNGP Basis Reference(s):
1. C.5.1-1100 (RPV CONTROL) 2. Ops Man B.01.01-03 (REACTOR AND VESSEL ASSEMBLY -INSTRUMENTATION AND CONTROLS)
3. Ops Man C.3 (SHUTDOWN PROCEDURE)
4. 9040 (TEMPORARY VESSEL LEVEL INSTRUMENTATION INSTALLATION AND RESTORATION)
5. Tech Spec Table 1.1-1 (MODES) 61 CU4 ECL: Notification of Unusual Event Initiating Condition:
Loss of Vital DC power for 15 minutes or longer. Operating Mode Applicability:
Cold Shutdown, Refueling Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
CU4.1 Indicated voltage is less than 110 VDC on required 125 VDC Vital DC buses for 15 minutes or longer. Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant. As used in this EAL, "required" means the Vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment.
For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train Bis in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.
The indicated voltage used in this threshold is based on battery sizing calculations.
The threshold is an average for both Division I and II batteries for battery voltages at 15 minutes prior to reaching the minimum required terminal voltage. The Division I and II-250 VDC battery systems need not be considered in this EAL because they supply power to large motor loads in the RCIC and HPCI systems and various non-critical loads. RCIC is an alternative source of make-up water for the reactor during normal plant shutdowns and transient events which lead to a loss of feedwater flow. HPCI is part of the Emergency Core Cooling System (ECCS) network. However, the Auto Depressurization System (ADS) is redundant in function to the HPCI system and does not require 250 VDC for operations.
Therefore, these systems need not be included in this EAL since loss of the 250 VDC battery systems would not cause core uncovering or loss of containment integrity.
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 CAl or CA3, or an IC in Recognition Category R. 62 MNGP Basis Reference(s):
1. USAR Section 8.5.1, Plant Electrical Systems -DC Power Supply Systems, Essential 250 Vdc System 2. USAR Section 8.5.2, Plant Electrical Systems-DC Power Supply Systems, 125 Vdc System 3. NE-36640-2, 125VDC Distribution Electrical Scheme 4. MNGP Calculation CA-02-179, 125 Volt Div. I Calculation
5. MNGP Calculation CA-02-192, 125 Volt Div. II Calculation
6. Technical Specification 3.8.5 (DC SOURCES-SHUTDOWN) 63 CU5 ECL: Notification of Unusual Event Initiating Condition:
Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability:
Cold Shutdown, Refueling, Defueled Emergency Action Levels: (CU5.l or CU5.2 or CU5.3) CU5.l Loss of ALL of the following onsite communication methods:
Commercial Telephones
Plant Telephones
Portable radios
Plant PA System CU5.2 Loss of ALL of the following Offsite Response Organization (ORO) communications methods:
Commercial Telephones
Direct Dedicated Telephones
Radio/Receiver Transmitter CU5.3 Loss of ALL of the following NRC communications methods:
Federal Telecommunications System (FTS)
Commercial Telephones Basis: This IC addresses a significant loss of on-site or offsite communications capabilities.
While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs andtheNRC.
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 infonnation via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). EAL CU5.l addresses a total loss of the communications methods used in support of routine plant operations.
EAL CU5.2 addresses a total loss of the communications methods used to notify all OR Os of an emergency declaration.
The OROs referred to here are the State of Minnesota, Wright County, and Sherburne County. EAL CU 5 .3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
64 MNGP Basis Reference(s):
1. USAR Section 10.3.8, Plant Auxiliary Systems -Plant Service Systems, Plant Communications System 2. MNGP Emergency Plan Section 7.2-Communication Systems 3. MNGP Emergency Plan Figure 13.7-Direct Dedicated Telephones (Hot Lines) 4. A.2-504 Emergency Communicator Duties in the TSC and OSC 65 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS UNUSUAL EVENT EUl Damage to a loaded cask CONFINEMENT BOUNDARY.
Op. Modes: All 66 EU1 ECL: Notification of Unusual Event Initiating Condition:
Damage to a loaded cask CONFINEMENT BOUNDARY.
Operating Mode Applicability:
All Emergency Action Levels: EUI.1 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than any of the values listed in Table El: Table El Location of Dose Rate Total Dose Rate (Neutron + Gamma mR/hr) HSM or HSM-H Front Surface 1400 HSM or HSM-H Door Centerline 200 End shield wall exterior 40 Basis: CONFINEMENT BOUNDARY:
The barrier(s) between areas containing spent fuel and the environment once the spent fuel is processed for dry storage. This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage. The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", which is also used in Recognition Category RIC RUl, is used here to distinguish between emergency and emergency conditions.
The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask. Security-related events for ISFSis are covered under I Cs HUI and HAI. MNGP Basis Reference(s):
1. 9508 (DSC TRANSFER FROM TRANSFER CASK TO HSM) 2. ISFSI Tech Spec 1.2.7f, HSM or HSM-H Dose Rates with a loaded Type 1 61BTH DSC Amendment 10 67 9 FISSION PRODUCT BARRIER ICS/EALS Recognition Category "F" Initiating Condition Matrix GENERAL EMERGENCY Loss of any two barriers and Loss or Potential Loss of the third barrier. FGl Op. Modes: Power Operation, Startup, Hot Shutdown SITE AREA EMERGENCY Loss or Potential Loss of any two barriers.
FSl Op. Modes: Power Operation, Startup, Hot Shutdown ALERT Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier. FAl Op. Modes: Power Operation, Startup, Hot Shutdown LOSS POTENTIAL LOSS FUEL CLAD LOSS POTENTIAL LOSS FUEL CLAD LOSS FUEL CLAD 68 LOSS POTENTIAL LOSS RCS POTENTIAL LOSS YES E.GL -Loss of ANY Two Bmriers AIDl Loss or Potential Loss of Third B:irrier LOSS POTENTIAL LOSS CONTAINMENT .ES.1-Loss or Potcntfol Loss of ANY Two Barriers '--------------_.i.EA.1-ANY Loss or ANY Potential Loss of EITHER Fuel Clad QB.RCS Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FGlGENERALEMERGENCY FSl SITE AREA EMERGENCY FAlALERT Loss of any two barriers and Loss or Loss or Potential Loss of any two barriers.
Any Loss or any Potential Loss of either the Potential Loss of the third barrier. Fuel Clad or RCS barrier. :,{}:t, n
'>;;.:,, ** * * *** ***** * * * * ,,, * * ,;{'::* ,:: .r ) **,'. *, '" ..*.
'. :*} * /
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/4.. '"' '*"' *. : : :: , ' ' ' ' ' .*,... . . . : . :* ' . .::.:t " . ' t"' ; ' ; ,. LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 1. RCS Activity 1. Primary Containment Pressure 1. Primary Containment Conditions A. Coolant activity is Not Applicable A. Primary Not Applicable A. UNPLANNED A. Primary greater than 300 containment rapid drop in containment
µCi/gm dose pressure greater primary pressure greater equivalent I-131. than 1.84 psig due containment than 56 psig to RCS leakage. pressure following OR primary B. Greater than or containment e.qual to 6% pressure rise hydrogen and OR greater than or B. Primary equal to 5% containment oxygen m pressure response Drywell or Torus not consistent with OR LOCA conditions.
c. HCL exceeded.
2. RPV Water Level 2. RPV Water Level 2. RPV Water Level A. SAMO entry is A. RPV water level A. RPV water level Not Applicable Not Applicable A. SAMO entry is required.
cannot be restored cannot be restored required.
and maintained and maintained above -126 in. or above -126 in. or cannot be cannot be determined.
determined.
3. Not Applicable
3. RCS Leak Rate 3. Primary Containment Isolation Failure Not Applicable Not Applicable A. UNISOLABLE A. UNISOLABLE A. UNISOLABLE Not Applicable break in ANY of the primary system direct downstream 69 LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS following:
MSL; HPCI;RWCU; RCIC as indicated by high flow/temperature isolation setpoints OR B. Emergency RPV Depressurization.
70 leakage that results in exceeding Control Room indication of EITHER of the following:
1. Max Normal Operating Temperature OR 2. Max Normal Operating Area Radiation Level. LOSS pathway to the environment exists after primary containment isolation signal OR B. Intentional pnmary containment venting per EOPs OR C. UNISOLABLE primary system leakage that results in exceeding Control Room indication of EITHER of the following:
1. Max Safe Operating Temperature.
OR 2. Max Safe Operating Area Radiation Level. POTENTIAL LOSS i :?,-*if'.'

"(**** * *: ::*_;:
...
LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 4. Primary Containment Radiation
4. Primary Containment Radiation
4. Primary Containment Radiation A. Containment High Not Applicable A. Containment High Not Applicable Not Applicable A. Containment High RangeRad RangeRad RangeRad (Drywell (Drywell Radiation) (Drywell Radiation) monitor reading Radiation) monitor reading greater than monitor reading greater than 6.2 E+Ol R/hr greater than 1.5 E+03 R/hr 3 .3 E +04 R/hr 5. Emergency Director Judgment 5. Emergency Director Judgment 5. Emergency Director Judgment A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the Emergency Emergency Emergency Director Emergency Emergency Emergency Director that Director that that indicates Loss Director that Director that Director that indicates Loss of indicates Potential of the RCS Barrier. indicates indicates Loss of indicates Potential the Fuel Clad Loss of the Fuel Potential Loss of the Containment Loss of the Barrier. Clad Barrier. the RCS Barrier. Barrier. Containment Barrier. 71 Basis Information For BWR EAL Fission Product Barrier Table FUEL CLAD BARRIER THRESHOLDS:
The Fuel Clad barrier consists of the zircalloy fuel bundle tubes that contain the fuel pellets. 1. RCS Activity Loss l.A This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. 2. There is no Potential Loss threshold associated with RCS Activity.
RPV Water Level Loss 2.A The Loss threshold represents any EOP requirement for entry into Severe Accident Management Guidelines.
This is identified in the BWR EPGs/SAGs when adequate core cooling cannot be assured. Potential Loss 2.A This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling. The RPV water level threshold is the same as RCS barrier Loss threshold 2.A. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency.
This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization.
EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration ofRPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.
72 The term "cannot be restored and maintained above" means the value ofRPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL).
Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority.
For such events, ICs SA5 or SS5 will dictate the need for emergency classification.
3. Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier, a potential loss of the fuel clad barrier is specified.
Not Applicable (included for numbering consistency between barrier tables) 4. Primary Containment Radiation Loss 4.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold 4.A since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency.
There is no Potential Loss threshold associated with Primary Containment Radiation.
5. Emergency Director Judgment Loss 5.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad Barrier is lost. Potential Loss 5 .A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is potentially lost. The Emergency Director 73 should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
74 RCS BARRIER THRESHOLDS:
The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping up to and including the isolation valves. 1. Primary Containment Pressure Loss I.A The 1.84 psig primary containment pressure is the drywell high pressure setpoint which indicates a LOCA by automatically initiating the ECCS or equivalent makeup system. There is no Potential Loss threshold associated with Primary Containment Pressure.
2. RPV Water Level Loss 2.A This water level corresponds to the top of active fuel and is used in the EOPs to indicate challenge to core cooling. The RPV water level threshold is the same as Fuel Clad barrier Potential Loss threshold 2.A. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency.
This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration ofRPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.
The term, "cannot be restored and maintained above," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The detennination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
75 In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL).
Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority.
For such events, ICs SAS or SSS will dictate the need for emergency classification.
There is no RCS Potential Loss threshold associated with RPV Water Level. 3. RCS Leak Rate Loss Threshold 3 .A Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated.
If it is determined that the ruptured line cannot be promptly isolated from the Control Room, the RCS barrier Loss threshold is met. Loss Threshold 3.B Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs). Even though the RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.
Potential Loss Threshold 3.A Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment.
A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.
The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification.
A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. An UNISOLABLE leak which is indicated by Max Normal Operating values escalates to a Site Area Emergency when combined with Containment Barrier Loss threshold 3.A (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.
76
4. Primary Containment Radiation Loss 4.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 4.A since it indicates a loss of the RCS Barrier only. There is no Potential Loss threshold associated with Primary Containment Radiation.
5. Emergency Director Judgment Loss 5.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost. Potential Loss 5.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
77 CONTAINMENT BARRIER THRESHOLDS:
The Primary Containment Barrier includes the drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
1. Primary Containment Conditions Loss I .A and I .B Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of primary containment integrity.
Primary containment pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity.
These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned.
The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition.
Potential Loss I .A The threshold pressure is the primary containment internal design pressure.
Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure.
A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier. Potential Loss l.B If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. The existence of an explosive mixture (2: 6% H2 and 2: 5% 02) means, at a minimum, that the containment hydrogen concentration is sufficient to support a hydrogen burn. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur. Potential Loss l.C The Heat Capacity Limit (HCL) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:
Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR 78
Suppression chamber pressure above Primary Containment Pressure Limit A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent. The HCL is a function of RPV pressure, suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment.
2. RPV Water Level There is no Loss threshold associated with RPV Water Level. Potential Loss 2.A The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold 2.A. The Potential Loss requirement for entry into the Severe Accident Management Guidelines indicates adequate core cooling cannot be assured and that core damage is possible.
BWR EPGs/SAGs specify the conditions when the EPGs are exited and SAMGs are entered. Entry into SAMGs is a logical escalation in response to the inability to assure adequate core cooling. PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increased potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency.
3. Primary Containment Isolation Failure These thresholds address incomplete containment isolation that allows an UNISOLABLE direct release to the environment.
A release path is 'direct' if it allows for the migration of radioactive material from the containment to the environment in a generally uninterrupted manner (e.g., little or no holdup time); therefore, within the context of a Containment Barrier Loss or Potential Loss threshold, a release path through the wetwell is a direct release path. Loss 3.A The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS). The existence of a filter is not considered in the threshold assessment.
Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream. Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components.
Minor 79 releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R ICs. Loss 3.B EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded.
Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed.
Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment.
In this situation conditions and trends are such that the Control Room staff has made a decision to perform an intentional controlled venting of the containment.
This intentional venting action results in a bypass of the primary containment, whether it is anticipatory or otherwise.
Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition.
Loss 3.C The Max Safe Operating Temperature and the Max Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded.
EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.
The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. In combination with RCS potential loss 3.A this threshold would result in a Site Area Emergency.
There is no Potential Loss threshold associated with Primary Containment Isolation Failure. 4. Primary Containment Radiation There is no Loss threshold associated with Primary Containment Radiation.
Potential Loss 4.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.
80 NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency.
5. Emergency Director Judgment Loss 5.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Containment barrier is lost. Potential Loss 5.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
MNGP Basis Reference(s):
1. Ops Man B.02.04 (MAIN STEAM) 2. Ops Man C.4-B.04.01.A (PRIMARY CONTAINMENT ISOLATION
-GROUP 1) 3. Ops Man C.6-005-A-25 (MAIN STEAM LINE HI FLOW CH A) 4. Ops Man C.6-005-A-26 (MAIN STEAM LINE HI FLOW CH B) 5. Ops Man C.6-003-B-56 (HIGH AREA TEMP STEAM LEAK) 6. Ops Man C.5.1-1100 (RPV CONTROL) 7. Ops Man C.5.1-1200 (PRIMARY CONTAINMENT CONTROL) 8. Ops Man C.5.1-1300 (SECONDARY CONTAINMENT CONTROL) 9. Ops Man C.5.1-2006 (RPV FLOODING)
10. Ops Man C.5.1-2007 (FAILURE TO SCRAM) 11. C.5-3505 (VENTING PRIMARY CONTAINMENT)
12. MNGP Calculation CA-04-194, Containment High Range Radiation Monitor (CHRRM) Response to Drywell Activity 13. USAR Section 5.2, Containment System-Primary Containment System 81
14. USAR Table Section 7.5, Plant Instrumentation and Control Systems-Plant Radiation Monitoring Systems 15. Tech Spec Table 3.3.3.1-1 (Post Accident Monitoring Instrumentation)
16. Tech Spec 3.6.1.1 (Primary Containment)
17. Tech Spec3.6.4.l (Secondary Containment)
18. NX-7831-197-1, Reactor Vessel & Internals 82 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS GENERAL SITE AREA ALERT UNUSUAL EVENT EMERGENCY EMERGENCY HGl HOSTILE HSl HOSTILE HAl HOSTILE HUl Confirmed ACTION resulting in ACTION within the ACTION within the SECURITY loss of physical control Plant PROTECTED OWNER CONDITION or threat. of the facility.
AREA. CONTROLLED AREA Op. Modes: All Op. Modes: All Op. Modes: All or airborne attack threat within 30 minutes. Op. Modes: All HU2 Seismic event greater than OBE levels. Op. Modes: All HU3 Hazardous event. Op. Modes: All HU4 FIRE potentially degrading the level of safety of the plant. Op. Modes: All HAS Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.
Op. Modes: All HS6 Inability to HA6 Control Room control a key safety evacuation resulting in function from outside transfer of plant control the Control Room. to alternate locations.
Op. Modes: All Op. Modes: All HG7 Other conditions HS7 Other conditions HA7 Other conditions HU7 Other conditions exist which in the exist which in the exist which in the exist which in the judgment of the judgment of the judgment of the judgment of the Emergency Director Emergency Director Emergency Director Emergency Director warrant declaration of a warrant declaration of a warrant declaration of warrant declaration of a General Emergency.
Site Area Emergency.
an Alert. (NO)UE. Op. Modes: All Op. Modes: All Op. Modes: All Op. Modes: All 83 HG1 ECL: General Emergency Initiating Condition:
HOSTILE ACTION resulting in loss of physical control of the facility.
Operating Mode Applicability:
All Emergency Action Levels: HGI.1 a. A HOSTILE ACTION is occurring or has occurred within the Plant PROTECTED AREA as reported by the Security Shift Supervisor.
AND b. EITHER of the following has occurred:
1. ANY of the following safety functions cannot be controlled or maintained.
Reactivity control
RPV water level
RCS heat removal OR 2. Damage to spent fuel has occurred or is IMMINENT.
Basis: HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). IMMINENT:
The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. PROTECTED AREA: The area surrounding the plant encompassed by the chain link fence and certain structures as defined in the Security Plan; excludes the ISFSI Protected Area. In areas where two fences are present, the inner fence is designated as the Protected Area barrier. This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions.
It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and 84 Independent Spent Fuel Storage Installation Security Program}.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.
This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
Security-sensitive information should be contained in non-public documents such as the Security Plan. MNGP Basis Reference(s):
1. MNGP Safeguards Contingency Plan 2. C.4-L (RESPONSE TO SECURITY THREATS) 3. NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan (and Independent Spent Fuel Storage Installation Security Program) 4. ND-95209, Monticello Main Plant Structures
5. QFl 775 (DEFINITIONS) 85 HG7 ECL: General Emergency Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency.
Operating Mode Applicability:
All Emergency Action Levels: HG7.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. Basis: HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). IMMINENT:
The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a General Emergency.
MNGP Basis Reference(s):
1. QFl 775 (DEFINITIONS) 86 HS1 ECL: Site Area Emergency Initiating Condition:
HOSTILE ACTION within the Plant PROTECTED AREA. Operating Mode Applicability:
All Emergency Action Levels: HS 1.1 A HOSTILE ACTION is occurring or has occurred within the Plant PROTECTED AREA as reported by the Security Shift Supervisor.
Basis: HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). PROTECTED AREA: The area surrounding the plant encompassed by the chain link fence and certain structures as defined in the Security Plan; excludes the ISFSI Protected Area. In areas where two fences are present, the inner fence is designated as the Protected Area barrier. This IC addresses the occurrence of a HOSTILE ACTION within the Plant PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering).
The Site Area Emergency declaration will mobilize ORO resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.
This IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA located outside the plant PROTECTED AREA; such an attack should be assessed using IC HAI. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. 87 Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.
This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
Security-sensitive information should be contained in non-public documents such as the Security Plan. Escalation of the emergency classification level would be via IC HGl. MNGP Basis Reference(s):
1. MNGP Safeguards Contingency Plan 2. C.4-L (RESPONSE TO SECURITY THREATS) 3. NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan (and Independent Spent Fuel Storage Installation Security Program) 4. ND-95209, Monticello Main Plant Structures
5. QFl 775 (DEFINITIONS) 88 HS6 ECL: Site Area Emergency Initiating Condition:
Inability to control a key safety function from outside the Control Room. Operating Mode Applicability:
Power Operation, Startup, Hot Shutdown, Cold Shutdown, Refueling Emergency Action Levels: [ Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 10 minutes has been exceeded, or will likely be exceeded.
HS6.1 a. An event has resulted in plant control being transferred from the Control Room to the alternate shutdown panel. Basis: AND b. Control of ANY of the following key safety functions is not reestablished within 10 minutes.
Reactivity control (Modes 1 and 2 only)
RPV water level
RCS heat removal This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time. The determination of whether or not "control" is established at the alternate shutdown panel is based on Emergency Director judgment.
The Emergency Director is expected to make a reasonable, informed judgment within 10 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).
Ops Man C.4-C, Shutdown Outside the Control Room, provides specific instructions for evacuating the Control Room and establishing plant control at the alternate shutdown panel. It should be noted here that analysis has shown that under worst case conditions (Assumed power level of 2004 MWt, Control Room fire coincident with loss of offsite power and reactor isolation and no available RPV injection) that indicated reactor level will decrease below the Top of Active Fuel (TAF) in approximately 11 minutes. Additionally, spurious operation of an SRV as a result of the fire event may lead to indicated reactor water level already being below T AF at the time of Operator arrival to the ASDS panel at the 10 minute mark. The EOPs would normally require depressurization before level reaches TAF however, thermal hydraulic analysis was performed and, given the above mentioned scenarios, conservatively assumed reactor manual depressurization to occur at 17 minutes. Escalation of the emergency classification level would be via IC FG 1 or CG 1. 89 MNGP Basis Reference(s):
1. Ops Man C.4-C (SHUTDOWN OUTSIDE THE CONTROL ROOM) 90 HS7 ECL: Site Area Emergency Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency.
Operating Mode Applicability:
All Emergency Action Levels: HS7 .1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.
Basis: HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a Site Area Emergency.
MNGP Basis Reference(s):
1. QFl 775 (DEFINITIONS) 91 HA1 ECL: Alert Initiating Condition:
HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes. Operating Mode Applicability:
All . Emergency Action Levels: (HAI. I or HAI .2) HAI .1 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervisor.
HAI .2 A validated notification from NRC of an aircraft attack threat within 30 minutes of the site. Basis: HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). OWNER CONTROLLED AREA: The OCA boundaries consist of the plant property enclosed by a three strand barbed wire fence and a posted boundary on the Wright County side of the nver. This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the Plant PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-I2, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program}.
As time and conditions allow, these events require a heightened state ofreadiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering).
The Alert declaration will also heighten the awareness of Off site Response Organizations, allowing them to be better prepared should it be necessary to consider further actions. This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. 92 Reporting of these types of events is adequately addressed by other EALs, or the requirements of IO CFR § 73.7I or IO CFR § 50.72. EAL HAI. I is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA. EAL HAI .2 addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state ofreadiness.
This EAL is met when the threat-related information has been validated in accordance with site procedures.
The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.
The status and size of the plane may be provided by NORAD through the NRC. In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.
This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
Security-sensitive information should be contained in non-public documents such as the Security Plan. Escalation of the emergency classification level would be via IC HS I. MNGP Basis Reference(s):
I. MNGP Safeguards Contingency Plan 2. C.4-L (RESPONSE TO SECURITY THREATS) 3. NEI 03-I2, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan (and Independent Spent Fuel Storage Installation Security Program) 4. FP-S-FSIP-08 (CONTINGENCY PLAN IMPLEMENTING PROCEDURES)
5. QFI 775 (DEFINITIONS) 93 HAS ECL: Alert Initiating Condition: Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability:
All Emergency Action Levels: Note: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.
HA5.l a. AND Release of a toxic, corrosive, asphyxiant or :flammable gas into any of the Table Hl plant rooms or areas: Table Hl Building Rooms Applicable Mode(s) Reactor Building All All Turbine Building All All Intake Structure All All b. Entry into the room or area is prohibited or impeded. Basis: This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown.
This condition represents an actual or potential substantial degradation of the level of safety of the plant. An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release. Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
An emergency declaration is not warranted if any of the following conditions apply.
The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the 94 gaseous release).
For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).
The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action. An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment.
This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death. This EAL does not apply to firefighting activities that automatically or manually activate a fire suppression system in an area, or to intentional inerting of containment).
Escalation of the emergency classification level would be via Recognition Category R, C or F I Cs. MNGP Basis Reference(s):
1. Ops Man C.l (STARTUP PROCEDURE)
2. Ops Man C.3 (SHUTDOWN PROCEDURE)
3. USAR Table*I.5-1, Location of High Energy Systems and Safe Shutdown Equipment by Volume 4. USAR Section 10.3, Plant Auxiliary Systems -Plant Service Systems 5. USAR Section 12.2, Plant Structures and Shielding-Plant Principal Structures and Foundations
6. ND-95209, Monticello Main Plant Structures 95 HAG ECL: Alert Initiating Condition:
Control Room evacuation resulting in transfer of plant control to alternate locations.
Operating Mode Applicability:
All Emergency Action Levels: HA6.1 An event has resulted in plant control being transferred from the Control Room to the alternate shutdown panel. Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety. Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations.
The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel.
Activation of the ERO and emergency response facilities will assist in responding to these challenges.
Escalation of the emergency classification level would be via IC HS6. MNGP Basis Reference(s):
1. Ops Man C.4-C (SHUTDOWN OUTSIDE THE CONTROL ROOM) 96 HA7 ECL: Alert Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert. Operating Mode Applicability:
All Emergency Action Levels: HA7.1 Basis: Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an Alert. MNGP Basis Reference(s):
1. QFI 775 (DEFINITIONS) 97 HU1 ECL: Notification of Unusual Event Initiating Condition:
Confirmed SECURITY CONDITION or threat. Operating Mode Applicability:
All Emergency Action Levels: (HUl. I or HUI .2 or HUI .3) HUI.I A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervisor.
HUl.2 Notification of a credible security threat directed at MNGP. HUI.3 A validated notification from the NRC providing information of an aircraft threat. Basis: HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). SECURITY CONDITION:
Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION. This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of IO CFR § 73.7I or IO CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAI, HSI and HGl. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and OROs. Security plans and terminology are based on the guidance provided by NEI 03-I2, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
EAL HUI. I references Security Shift Supervisor because this is the individual trained to confirm that a security event is occurring or has occurred.
Training on security event confirmation and classification is controlled dueto the nature of Safeguards and I 0 CFR § 2.39 information.
EAL HUI .2 addresses the receipt of a credible security threat. The credibility of the threat is 98 assessed in accordance with site procedures.
EAL HUl .3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee ifthe threat involves an aircraft.
The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with site procedures.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.
This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
Security-sensitive information should be contained in non-public documents such as the Security Plan. Escalation of the emergency classification level would be via IC HAl. MNGP Basis Reference(s):
1. MNGP Safeguards Contingency Plan 2. C.4-L (RESPONSE TO SECURITY THREATS) 3. NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan (and Independent Spent Fuel Storage Installation Security Program) 4. FP-S-FSIP-08 (CONTINGENCY PLAN IMPLEMENTING PROCEDURES)
5. QFl 775 (DEFINITIONS) 99 ECL: Notification of Unusual Event Initiating Condition:
Seismic event greater than OBE levels. Operating Mode Applicability:
All Emergency Action Levels: HU2.l Seismic event greater than Operating Basis Earthquake (OBE) as indicated by Annunciator OPERATIONAL BASIS EARTHQUAKE (6-C-13) received.
Basis: HU2 This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections).
Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant. The Control Room annunciator EARTHQUAKE (6-C-08) alarms either by the seismic trigger of the Accelerograph Recording System or seismic switch of the Seismic Annunciator System. The annunciator OPERATIONAL BASIS EARTHQUAKE (6-C-13) alarms when its switch senses an acceleration 2: 0.03g. The Accelerograph Recording System records accelerations in three directions, longitudinal, transversal and vertical.
This IC is based on the USAR operating basis earthquake (OBE) of 0.04g vertical or 0.06g horizontal.
Classification for this IC is to occur upon receipt of annunciator 6-C-13 as it is immediately available to CR personnel and is readily assessed.
Seismic events of this magnitude (i.e.,> OBE but::::; DBE) have been analyzed and designed for at the MNGP. However, events of this magnitude can result in plant equipment being subjected to forces that require further engineering attention.
Event verification with external sources should not be necessary during or following an OBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., typical lateral accelerations are in excess of 0.08g). The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9. 100 MNGP Basis Reference(s):
1. Ops Man A.6 (ACTS OF NATURE) 2. Ops Man B.05.14-01 (SEISMIC MONITORING-SYSTEM OPERATION)
3. C.4-B.05.14.A (EARTHQUAKE)
4. C.6-006-C-08 (EARTHQUAKE)
5. C.6-006-C-13 (OPERATIONAL BASIS EARTHQUAKE)
6. USAR Section 2.6.5, Seismic Monitoring System 7. USAR Section 12.2.1, Plant Principal Structures and Foundations, Design Basis 101 HU3 ECL: Notification of Unusual Event Initiating Condition:
Hazardous event. Operating Mode Applicability:
All Emergency Action Levels: (HU3.l or HU3.2 or HU3.3 or HU3.4 or HU3.5 or HU3.6) Note: EAL HU3.4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
HU3.l A tornado strike within the Plant PROTECTED AREA. HU3.2 Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode. HU3 .3 Movement of personnel within the Plant PROTECTED AREA is impeded due to an off site event involving hazardous materials (e.g., an off site chemical spill or toxic gas release).
HU3.4 A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.
HU3.5 River level greater than 918 ft elevation.
HU3.6 River level less than 902.4 ft elevation.
Basis: PROTECTED AREA: The area surrounding the plant encompassed by the chain link fence and certain structures as defined in the Security Plan; excludes the ISFSI Protected Area. ill areas where two fences are present, the inner fence is designated as the Protected Area barrier. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. EAL HU3.1 addresses a tornado striking (touching down) within the Protected Area. EAL HU3.2 addresses flooding of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.
Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. 102 EAL HU3.3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the Plant PROTECTED AREA. EAL HU3.4 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles.
Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road. This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. EAL HU3 .5 addresses a potential flood condition.
The 918 ft elevation is selected for this EAL because it is the first elevation at which procedure actions are required to address flooding situations.
EAL HU3.6 addresses low river flow conditions.
The low river water level threshold (902.4 ft elevation) corresponds to the low river flow threshold of 240 CFS. Low river level (i.e., flow) may be a precursor to loss of the ultimate heat sink and warrants further management attention.
Escalation of the emergency classification level would be based on I Cs in Recognition Categories A, F, S or C. MNGP Basis Reference(s):
l. OpsManA.6 (ACTS OF NATURE) 2. USAR Section 10.3, Plant Auxiliary Systems -Plant Service Systems 3. USAR Section 12.2, Plant Structures and Shielding-Plant Principal Structures and Foundations
4. USAR Appendix G, Chapter 3, Probable Maximum Flood Determination
5. ND-95208, Monticello Property Map 6. ND-95209, Monticello Main Plant Structures 103 r HU4 ECL: Notification of Unusual Event Initiating Condition:
FIRE potentially degrading the level of safety of the plant. Operating Mode Applicability:
All Emergency Action Levels: (HU4.1 or HU4.2 or HU4.3 or HU4.4) Note: The Emergency Director should declare the Unusual Event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
HU4.l a. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications:
Report from the field (i.e., visual observation)
Receipt of multiple (more than 1) fire alarms or indications
Field verification of a single fire alarm AND b. . The FIRE is located within ANY of the Table H2 plant rooms or areas. HU4.2 a. Receipt of a single fire alarm (i.e., no other indications of a FIRE). AND b. The FIRE is located within ANY of the Table H2 plant rooms or areas. AND c. The existence of a FIRE is not verified within 30-minutes of alarm receipt. HU4.3 A FIRE within the Plant PROTECTED AREA or ISFSI PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication.
HU4.4 A FIRE within the Plant PROTECTED AREA or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
TableH2 Building Name Room(s)/Area(s) with Safety Equipment Reactor Building All HPCI Building All Turbine Building All Control and Administration Control Room, Cable Spreading Room, and Building Battery Rooms Diesel Generator Building All Diesel Fuel Oil Transfer House All EFT Building All Intake Structure All Basis: 104
,-----FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
ISFSI (INDEPENDENT SPENT FUEL STORAGE INSTALLATION):
A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. ISFSI PROTECTED AREA: The area surrounding the Independent Spent Fuel Storage Installation encompassed by the double chain link fence surrounding the ISFSI as defined in the Security Plan; the ISFSI Protected Area is excluded from the Plant Protected Area. PROTECTED AREA: The area surrounding the plant encompassed by the chain link fence and certain structures as defined in the Security Plan; excludes the ISFSI Protected Area. In areas where two fences are present, the inner fence is designated as the Protected Area barrier. This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. EALHU4.1 The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. Upon receipt, operators will take prompt actions to confinn the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed.
Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. EALHU4.2 This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.
A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.
If an actual FIRE is verified by a report from the field, then EAL HU4. l is immediately applicable, and the emergency must be declared ifthe FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.
105 EALHU4.3 In addition to a FIRE addressed by EAL HU4.1 or EAL HU4.2, a FIRE within the Plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA. EALHU4.4 If a FIRE within the Plant PROTECTED AREA or ISFSI PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded.
The dispatch of an off site firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish.
Declaration is not necessary if the agency resources are placed on stand-by, or supporting extinguishment recovery or investigation actions. Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part: Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.
Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.
In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in EAL HU4.2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9. MNGP Basis Reference(s):
1. USAR Section 10.3, Plant Auxiliary Systems -Plant Service Systems 2. USAR Section 12.2, Plant Structures and Shielding-Plant Principal Structures and Foundations 106
3. USAR Appendix I, Table I.5-1, Location of High Energy System and Safe Shutdown Equipment by Volume 4. USAR Appendix J.4, Fire Protection Program -Safe Shutdown Analysis 5. ND-95209, Monticello Main Plant Structures
6. NF-36300-1-2, Block Wall Schedule Reactor Building 7. NF-36300-1-3, Block Wall Schedule Turbine Building 8. NF-3600-1-4, Block Wall Schedule Plant Admin Building and Offgas Stack 9. 4 A WI-01.03.01 (QUALITY ASSURANCE PROGRAM BOUNDARY)
10. Ops Man B.08.05-05 (FIRE PROTECTION SYSTEM OPERATION) 107 HU7 ECL: Notification of Unusual Event Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a NUE. Operating Mode Applicability:
All Emergency Action Levels: HU7 .1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a NUE. MNGP Basis Reference(s):
1. QFl 775 (DEFINITIONS) 108 11 SYSTEM MALFUNCTION ICS/EALs GENERAL EMERGENCY SGl Prolonged loss of all offsite and all onsite AC power to essential buses. Op. Modes: Power Operation, Startup, , Hot Shutdown SITE AREA EMERGENCY SSl Loss of all offsite and all onsite AC power to essential buses for 15 minutes or longer. Op. Modes: Power Operation, Startup, , Hot Shutdown SSS Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal. Op. Modes: Power Operation 109 ALERT UNUSUAL EVENT SAl Loss of all but one SUl Loss of all offsite AC power source to AC power capability to essential buses for 15 essential buses for 15 minutes or longer. Op. Modes: Power Operation, Startup, , Hot Shutdown SA2 UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.
Op. Modes: Power Operation, Startup, , Hot Shutdown SAS Automatic or manual scram fails to shutdown the reactor, and subsequent manual actions taken at the main control boards are not successful in shutting down the reactor. Op. Modes: Power Operation minutes or longer. Op. Modes: Power Operation, Startup, , Hot Shutdown SU2 UNPLANNED loss of Control Room indications for 15 minutes or longer. Op. Modes: Power Operation, Startup, , Hot Shutdown SU3 Reactor coolant activity greater than Technical Specification allowable limits. Op. Modes: Power Operation, Startup, , Hot Shutdown SU4 RCS leakage for 15 minutes or longer. Op. Modes: Power Operation, Startup, , Hot Shutdown SUS Automatic or manual scram fails to shutdown the reactor. Op. Modes: Power Operation GENERAL SITE AREA ALERT UNUSUAL EVENT EMERGENCY EMERGENCY SU6 Loss of all onsite or offsite communications capabilities.
Op. Modes: Power Operation, Startup, , Hot Shutdown SGS Loss of all AC SSS Loss of all Vital and Vital DC power DC power for 15 minutes sources for 15 minutes or or longer. longer. Op. Modes: Power Op. Modes: Power Operation, Startup, , Hot Operation, Startup, , Hot Shutdown Shutdown SA9 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Op. Modes: Power Operation, Startup, , Hot Shutdown 110 SG1 ECL: General Emergency Initiating Condition:
Prolonged loss of all offsite and all onsite AC power to essential buses. Operating Mode Applicability:
Power Operation, Startup, Hot Shutdown Emergency Action Levels: Note: The Emergency Director should declare the General Emergency promptly upon determining that 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months has been exceeded, or will likely be exceeded.
'---------
SGl.1 a. Loss of ALL offsite and ALL onsite AC power to essential buses 15 and 16. Basis: AND b. EITHER of the following:
Restoration of at least one AC essential bus in less than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is not likely.
Reactor vessel water level cannot be restored and maintained above -149" (Minimum Steam Cooling RPV Water Level) This IC addresses a prolonged loss of all power sources to AC essential buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers.
In addition, fission product barrier monitoring capabilities may be degraded under these conditions.
The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG 1. This will allow additional time for implementation of offsite protective actions. Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC essential bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.
The estimate for restoring at least one essential bus should be based on a realistic appraisal of the situation.
Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public. The EAL will also require a General Emergency declaration ifthe loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core. MNGP Basis Reference(s):
1. Ops Man B.09.06-01 (4.16 KV STATION AUXILIARY -FUNCTION AND GENERAL DESCRIPTION OF SYSTEM) 111
2. Ops Man C.4-B.09.02.A (STATION BLACKOUT)
3. Ops Man C.4-B.09.02.B (LOSS OF NORMAL OFF-SITE POWER) 4. Ops Man C.4-B.09.06.C (LOSS OF BUS 15 OR BUS 16) 5. C.5.1-1100 (RPV CONTROL) 6. USAR Section 8.2.1, Plant Electrical Systems-Transmission System, Network Interconnections
7. USAR Section 8.5.1.1, Plant Electrical Systems-DC Power Supply Systems, Essential 250 V de Systems, Design Basis 8. USAR Section 8.12, Plant Electrical Systems -Station Blackout 9. USAR Figure 8.4-1, Diesel Generation System One Line Diagram 10. NF-36175, Single Line Diagram-Station Connection
11. Tech Spec 3.8.l (AC SOURCES -OPERATING)
12. Tech Spec 3.8.7 (DISTRIBUTION SYSTEMS -OPERATING) 112 SGS ECL: General Emergency Initiating Condition:
Loss of all AC and Vital DC power sources for 15 minutes or longer. Operating Mode Applicability:
Power Operation, Startup, Hot Shutdown Emergency Action Levels: Note: The Emergency Director should declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
SG8. l a. Loss of ALL offsite and ALL onsite AC power to essential buses 15 and 16 for 15 minutes or longer. Basis: AND b. Indicated voltage is less than 110 VDC on ALL 125 VDC Vital DC buses for 15 minutes or longer. This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.
The indicated voltage used in this threshold is based on battery sizing calculations.
The threshold is an average for both Division I and II batteries for battery voltages at 15 minutes prior to reaching the minimum required terminal voltage. The Division I and II -250 VDC battery systems need not be considered in this EAL because they supply power to large motor loads in the RCIC and HPCI systems and various non-critical loads. RCIC is an alternative source of make-up water for the reactor during normal plant shutdowns and transient events which lead to a loss of feedwater flow. HPCI is part of the Emergency Core Cooling System (ECCS) network. However, the Auto Depressurization System (ADS) is redundant in function to the HPCI system and does not require 250 VDC for operations.
Therefore, these systems need not be included in this EAL since loss of the 250 VDC battery systems would not cause core uncovering or loss of containment integrity.
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. MNGP Basis Reference(s):
1. Ops Man B.09.06-01 (4.16 KV STATION AUXILIARY -FUNCTION AND GENERAL DESCRIPTION OF SYSTEM) 2. Ops Man C.4-B.09.02.A (STATION BLACKOUT)
3. Ops Man C.4-B.09.02.B (LOSS OF NORMAL OFF-SITE POWER) 113
4. Ops Man C.4-B.09.06.C (LOSS OF BUS 15 OR BUS 16) 5. USAR Section 8.2.1, Plant Electrical Systems -Transmission System, Network Interconnections
6. USAR Section 8.5.1.1, Plant Electrical Systems-DC Power Supply Systems, Essential 250 V de System, Design Basis 7. USAR Section 8.12, Plant Electrical Systems -Station Blackout 8. USAR Figure 8.4-1, Diesel Generation System One Line Diagram 9. NF-36175, Single Line Diagram -Station Connection
10. Tech Spec 3.8.l (AC SOURCES-OPERATING)
11. Tech Spec 3.8.4 (DC SOURCES -OPERATING)
12. Tech Spec 3.8.7 (DISTRIBUTION SYSTEMS-OPERATING)
13. MNGP Calculation CA-02-179, 125 Volt Div. I Calculation
14. MNGP Calculation CA-02-192, 125 Volt Div. II Calculation 114 ECL: Site Area Emergency Initiating Condition:
Loss of all offsite and all onsite AC power to essential buses for 15 minutes or longer. Operating Mode Applicability:
Power Operation, Startup, Hot Shutdown Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
551 SSl.1 Loss of ALL offsite and ALL onsite AC power to essential buses 15 and 16 for 15 minutes or longer. Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.
This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via ICs RGI, FGl or SGl. MNGP Basis Reference(s):
1. Ops Man B.09.06-01 (4.16 KV STATION AUXILIARY -FUNCTION AND GENERAL DESCRIPTION OF SYSTEM) 2. Ops Man C.4-B.09.02.A (STATION BLACKOUT)
3. Ops Man C.4-B.09.02.B (LOSS OF NORMAL OFF-SITE POWER) 4. Ops Man C.4-B.09.06.C (LOSS OF BUS 15 OR BUS 16) 5. USAR Section 8.2.1, Plant Electrical Systems-Transmission System, Network Interconnections
6. USAR Figure 8.4-1, Diesel Generation System One Line Drawing 7. NF-36175, Single Line Diagram-Station Connection
8. Tech Spec 3.8.1 (AC SOURCES-OPERATING)
9. Tech Spec 3.8.7 (DISTRIBUTION SYSTEMS-OPERATING) 115 SSS ECL: Site Area Emergency Initiating Condition:
Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal. Operating Mode Applicability:
Power Operation Emergency Action Levels: SS5.l a. An automatic or manual scram did not reduce reactor power to less than 4%. Basis: AND b. All manual actions to shutdown the reactor are not successful in reducing reactor power to less than 4%. AND c. EITHER of the following conditions exist:
Reactor vessel water level cannot be restored and maintained above -149" (Minimum Steam Cooling RPV Water Level)
Heat Capacity Limit (HCL) exceeded This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.
In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs.
This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency 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 RGI or FGl. MNGP Basis Reference(s):
1. C.5.1-1100 (RPV CONTROL) 2. C.5.1-1200 (PRIMARY CONTAINMENT CONTROL) 3. C.5.1-2007 (FAILURE TO SCRAM) 116
4. C.5-3101 (ALTERNATE ROD INSERTION)
5. C.4-A (REACTOR SCRAM) 6. USAR Table 7 .6-1, Typical Reactor Protection System Scram Setpoints
7. Tech Spec Table 3.3.1.1-1 (REACTOR PROTECTION SYSTEM INSTRUMENTATION) 117 ECL: Site Area Emergency Initiating Condition:
Loss of all Vital DC power for 15 minutes or longer. Operating Mode Applicability:
Power Operation, Startup, Hot Shutdown Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
558 SS8. l Indicated voltage is less than 110 VDC on ALL 125 VDC 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 SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public. The indicated voltage used in this threshold is based on battery sizing calculations.
The threshold is an average for both Division I and II batteries for battery voltages at 15 minutes prior to reaching the minimum required terminal voltage. The Division I and II-250 VDC battery systems need not be considered in this EAL because they supply power to large motor loads in the RCIC and HPCI systems and various non-critical loads. RCIC is an alternative source of make-up water for the reactor during normal plant shutdowns and transient events which lead to a loss of feedwater flow. HPCI is part of the Emergency Core Cooling System (ECCS) network. However, the Auto Depressurization System (ADS) is redundant in function to the HPCI system and does not require 250 VDC for operations.
Therefore, these systems need not be included in this EAL since loss of the 250 VDC battery systems would not cause core uncovering or loss of containment integrity.
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, FGl or SGS. MNGP Basis Reference(s):
1. USAR Section 8.5.1, Plant Electrical Systems -DC Power Supply Systems, Essential 250 Vdc System 2. USAR Section 8.5.2, Plant Electrical Systems-DC Power Supply Systems, 125 Vdc System 3. NE-36640-2, 125VDC Distribution Electrical Scheme 4. MNGP Calculation CA-02-179, 125 Volt Div. I Calculation
5. MNGP Calculation CA-02-192, 125 Volt Div. II Calculation 118
6. Tech Spec 3.8.4 (DC SOURCES -OPERATING) 119 SA1 ECL: Alert Initiating Condition:
Loss of all but one AC power source to essential buses for 15 minutes or longer. Operating Mode Applicability:
Power Operation, Startup, Hot Shutdown Emergency Action Levels: Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
SAl.1 Basis: a. AC power capability to essential buses 15 and 16 is reduced to a single power source (Table S 1) for 15 minutes or longer. AND b. Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS. Table Sl lR Reserve Transformer lAR Reserve Transformer 2R Auxiliary Transformer
11 Emergency Diesel Generator
12 Emergency Diesel Generator SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
This IC describes a significant degradation of offsite and onsite AC power sources (Table S 1) such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.
This IC provides an escalation path from IC SUI. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an essential bus. Some examples of this condition are presented below.
A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
A loss of emergency power sources (e.g., onsite diesel generators) with a single train of essential buses being back-fed from an 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 SSL 120 MNGP Basis Reference(s):
1. Ops Man B.09.06-01 (4.16 KV STATION AUXILIARY -FUNCTION AND GENERAL DESCRIPTION OF SYSTEM) 2. Ops Man C.4-B.09.02.A (STATION BLACKOUT)
3. Ops Man C.4-B.09.02.B (LOSS OF NORMAL OFF-SITE POWER) 4. Ops Man C.4-B.09.06.C (LOSS OF BUS 15 OR BUS 16) 5. USAR Section 8.2.1, Plant Electrical Systems-Transmission System, Network Interconnections
6. USAR Figure 8.4-1, Diesel Generation System One Line Diagram 7. NF-36175, Single Line Diagram -Station Connection
8. Tech Spec 3.8.1 (AC SOURCES-OPERATING)
9. Tech Spec 3.8.7 (DISTRIBUTION SYSTEMS-OPERATING) 121 SA2 ECL: Alert Initiating Condition:
UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.
Operating Mode Applicability:
Power Operation, Startup, Hot Shutdown Emergency Action Levels: Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
SA2.l a. AND An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer. Reactor Power RPV Water Level RPV Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature
b. ANY of the following transient events in progress.
Automatic or manual runback greater than 25% thermal reactor power
Electrical load rejection greater than 25% full electrical load
Reactor scram
ECCS actuation
Thermal power oscillations greater than 10% Basis: UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant. As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s
). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room. An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an 122 NRC event report is required.
The event would be reported if it significantly impaired the capability to perform emergency assessments.
In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.
In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level would be via ICs FSl or IC RSl. MNGP Basis Reference(s):
1. Ops Man C.4-B.05.13.A (LOSS OF ANNUNCIATOR)
2. Ops Man B.05.10 (PROCESS COMPUTER)
3. USAR Section 7.1, Plant Instrumentation and Control Systems-Summary Description
4. USAR Section 7.13, Plant Instrumentation and Control Systems -Safety Parameter Display System 123 SAS ECL: Alert Initiating Condition:
Automatic or manual scram fails to shutdown the reactor, and subsequent manual actions taken at the main control boards are not successful in shutting down the reactor. Operating Mode Applicability:
Power Operation Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
Emergency Action Levels: SA5.l a. An automatic or manual scram did not reduce reactor power to less than 4%. Basis: AND b. Manual actions taken at the main control boards are not successful in reducing reactor power to less than 4%. This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the main control boards (C-OS) to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the main control boards since this event entails a significant failure of the RPS. A manual action at the main control boards is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., see SUS). This action does not include manually driving in control rods or implementation of boron injection strategies.
If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the main control boards (e.g., locally opening breakers).
Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the main control boards". Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS5. Depending upon plant responses and symptoms, escalation is also possible via IC FSl. Absent the plant conditions needed to meet either IC SSS or FSl, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.
124 A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
The MNGP EOP entry condition for a failure to scram is defined to be a power above the APRM downscale setpoint (4%) following a reactor scram indicates that significant power is being generated.
If a scram is successful, reactor power will be indicated to be less than 4%. Therefore, reducing power to LESS THAN 4% is used in SAS.I as indication that a scram was successful.
MNGP Basis Reference(s):
I. C.4-A (REACTOR SCRAM) 2. C.5.1-1000 (EOP INTRODUCTION)
3. C.5.1-1100 (RPV CONTROL -FLOWCHART)
4. C.5.1-2007 (FAILURE TO SCRAM) 5. C.5-3101 (ALTERNATE ROD INSERTION)
6. USAR Table 7.6-1, Typical Reactor Protection System Scram Setpoints
7. Tech Spec Table 3.3.1.1-1 (REACTOR PROTECTION SYSTEM INSTRUMENTATION) 125 SA9 ECL: Alert Initiating Condition:
Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Operating Mode Applicability:
Power Operation, Startup, Hot Shutdown Emergency Action Levels: SA9.1 a. The occurrence of ANY of the following hazardous events: Basis: AND
Seismic event (Earthquake)
Internal or external flooding event
High winds or tornado strike
FIRE
EXPLOSION
River level greater than 919 ft el.
River level less than 900.5 ft el.
Other events with similar hazard characteristics as determined by the Shift Manager b. EITHER of the following:
Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.
The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode. EXPLOSION:
A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.
Such events may require a post-event inspection to determine if the attributes of an explosion are present. FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, 126 and therefore represents an actual or potential substantial degradation of the level of safety of the plant. The first threshold for EAL SA9 .1 b addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.
The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. The second threshold for EAL SA9.lb addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components.
Operators will make this determination based on the totality of available event and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC FSl or RSI. MNGP Basis Reference(s):
1. Ops Man A.6 (ACTS OF NATURE) 2. Ops Man B.05.14 (SEISMIC MONITORING SYSTEM) 3. Ops Man B.05.16-01 (METEOROLOGICAL MONITORING-FUNCTION & GENERAL DESCRIPTION OF SYSTEM) 4. Ops Man B.06.04 (CIRCULATING WATER SYSTEM) 5. C.4-B.05.14.A (EARTHQUAKE)
6. C.6-006-C-08 (EARTHQUAKE)
7. C.6-006-C-13 (OPERATIONAL BASIS EARTHQUAKE)
8. USAR Section 10.3, Plant Auxiliary Systems -Plant Service Systems 9. USAR Section 12.2, Plant Structures and Shielding, Plant Principal Structures and Foundations
10. USAR Appendix G, Chapter 3, Probable Maximum Flood Determination
11. USAR Table I.5-1, Location of High Energy Systems and Safe Shutdown Equipment by Volume 12. USAR Appendix J.4, Fire Protection Program-Safe Shutdown Analysis 13. ND-95208, Monticello Property Map 14. ND-95209, Monticello Main Plant Structures
15. 4 AWI-01.03.01 (QUALITY ASSURANCE PROGRAM BOUNDARY) 127 SU1 ECL: Notification of Unusual Event Initiating Condition:
Loss of all offsite AC power capability to essential buses for 15 minutes or longer. Operating Mode Applicability:
Power Operation, Startup, Hot Shutdown Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
SUI.I Loss of ALL offsite AC power capability (Table S2) to essential buses 15 and 16 for 15 minutes or longer. Table S2 1 R Reserve Transformer lAR Reserve Transformer 2R Auxiliary Transformer Basis: This IC addresses a prolonged loss of offsite power (Table S2). The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC essential buses. This condition represents a potential reduction in the level of safety of the plant. For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the essential 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 SAL MNGP Basis Reference(s):
1. Ops Man B.09.06-01 (4.16 KV STATION AUXILIARY -FUNCTION AND GENERAL DESCRIPTION OF SYSTEM) 2. Ops Man C.4-B.09.02.A (STATION BLACKOUT)
3. Ops Man C.4-B.09.02.B (LOSS OF NORMAL OFF-SITE POWER) 4. Ops Man C.4-B.09.06.C (LOSS OF BUS 15 OR BUS 16) 5. USAR Section 8.2.1, Plant Electrical Systems-Transmission System, Network Interconnections 128
6. USAR Figure 8.4-1, Diesel Generation System One Line Diagram 7. NF-36175, Single Line Diagram-Station Connection
8. Tech Spec 3.8.1 (AC SOURCES-OPERATING)
9. Tech Spec 3.8.7 (DISTRIBUTION SYSTEMS-OPERATING) 129 SU2 ECL: Notification of Unusual Event Initiating Condition:
UNPLANNED loss of Control Room indications for 15 minutes or longer. Operating Mode Applicability:
Power Operation, Startup, Hot Shutdown Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
SU2.1 Basis: An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer. Reactor Power RPV Water Level RPV Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant. As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s).
For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room. An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.
The event would be reported if it significantly impaired the capability to perform emergency assessments.
In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be 130 more significant than simply a reportable condition.
In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, ifthe value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level would be via IC SA2. MNGP Basis Reference(s):
1. Ops Man C.4-B.05.13.A (LOSS OF ANNUNCIATOR)
2. Ops Man B.05.10 (PROCESS COMPUTER)
3. USAR Section 7.1, Plant Instrumentation and Control Systems-Summary Description
4. USAR Section 7.13, Plant Instrumentation and Control Systems -Safety Parameter Display System 131 SU3 ECL: Notification of Unusual Event Initiating Condition:
Reactor coolant activity greater than Technical Specification allowable limits. Operating Mode Applicability:
Power Operation, Startup, Hot Shutdown Emergency Action Levels: (SU3.1 or SU3.2) SU3.l Offgas Pretreatment Radiation Monitor (RM-17-150A or RM-17-150B) high radiation alarm (4-A-12) received.
SU3.2 Coolant sample activity greater than 0.2 µCi/gm dose equivalent I-131. Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications.
This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. For EAL SU3.l the Offgas high radiation alarm received from RM-17-150A or RM-17-150B is set to meet the Technical Specification allowable limit of less than or equal to 2.6E+5 µCi/sec after decay of 30 minutes. EAL SU3.2 addresses reactor coolant samples exceeding coolant Technical Specifications.
Escalation of the emergency classification level would be via I Cs FAl or the Recognition Category R ICs. MNGP Basis Reference(s):
1. Ops Man B.07.02.02-01 (OFF-GAS HOLDUP SYSTEM-FUNCTION
& GENERAL DESCRIPTION OF SYSTEM) 2. C.6-004-A-12 (OFF-GAS HI RADIATION)
3. Tech Spec 3.4.6 (RCS SPECIFIC ACTIVITY)
4. Tech Spec 3.7.6 (MAIN CONDENSER OFFGAS) 5. Tech Spec 3.10.1 (INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION) 132 SU4 ECL: Notification of Unusual Event Initiating Condition:
RCS leakage for 15 minutes or longer. Operating Mode Applicability:
Power Operation, Startup, Hot Shutdown Emergency Action Levels: (SU4.l or SU4.2 or SU4.3) Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
SU4.1 SU4.2 SU4.3 Basis: RCS unidentified or pressure boundary leakage greater than 10 gpm for 15 minutes or longer. RCS identified leakage greater than 25 gpm for 15 minutes or longer. Leakage from the RCS to a location outside primary containment greater than 25 gpm for 15 minutes or longer. This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant. EAL SU4.1 and EAL SU4.2 are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications).
EAL SU4.3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs thus apply to leakage into the containment or a location outside of containment.
The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications.
Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation).
EAL SU4.1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage. The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification.
A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL. 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 I Cs of Recognition Category R or F. 133 MNGP Basis Reference(s):
I. Ops Man C.4-B.04.01.F (LEAK INSIDE PRIMARY CONTAINMENT)
2. Ops Man C.6-004-B-03 (DRYWELL SUMP VALVES CLOSED) 3. Ops Man C.6-004-B-17 (DRYWELL FLOOR DRAIN SUMP HI LEVEL) 4. Ops Man C.6-004-B-18 (DRYWELL EQUIP DRAIN LEAK RATE CHANGE HI) 5. Tech Spec 3.4.4 (RCS OPERATIONAL LEAKAGE) 134 SUS ECL: Notification of Unusual Event Initiating Condition:
Automatic or manual scram fails to shutdown the reactor. Operating Mode Applicability:
Power Operation Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
Emergency Action Levels: (SU5.1) SU5.l a. AND An initial automatic or manual scram did not reduce reactor power to less than 4%. b. ANY of the following is successful in reducing reactor power to less than 4%:
Manual scram pushbuttons
Mode switch to shutdown
Alternate rod insertion (ARI)
Subsequent automatic scram Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the main control boards (C-05) or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the main control boards to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the main control boards to shutdown the reactor (e.g., manual scram pushbuttons, mode switch to Shutdown, Alternate Rod Insertion (ARI)) using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the main control boards is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies.
Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the main control boards". 135 Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the main control boards are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SAS. Depending upon the plant response, escalation is also possible via IC F Al. Absent the plant conditions needed to meet either IC SAS or FAl, 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 scram* signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.
If the signal causes a plant transient that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.
If the signal does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.
MNGP Basis Reference(s):
1. C.4-A (REACTOR SCRAM) 2. C.S.1-1000 (EOP INTRODUCTION)
3. C.S.1-1100 (RPV CONTROL-FLOWCHART)
4. C.S.1-2007 (FAILURE TO SCRAM) S. C.S-3101 (ALTERNATE ROD INSERTION)
6. USAR Table 7 .6-1, Typical Reactor Protection System Scram Setpoints
7. Tech Spec Table 3.3.1.1-1 (REACTOR PROTECTION SYSTEM INSTRUMENTATION) 136 SU6 ECL: Notification of Unusual Event Initiating Condition:
Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability:
Power Operation, Startup, Hot Shutdown Emergency Action Levels: (SU6.l or SU6.2 or SU6.3) SU6.l Loss of ALL of the following onsite communication methods:
Commercial Telephones
Plant Telephones
Portable radios
Plant PA System SU6.2 Loss of ALL of the following Offsite Response Organization (ORO) communications methods:
Commercial Telephones
Direct Dedicated Telephones
Radio/Receiver Transmitter SU6.3 Loss of ALL of the following NRC communications methods:
Federal Telecommunications System (FTS)
Commercial Telephones Basis: This IC addresses a significant loss of on-site or offsite communications capabilities.
While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). EAL SU6.l addresses a total loss of the communications methods used in support ofroutine plant operations.
EAL SU6.2 addresses a total loss of the communications methods used to notify all OR Os of an emergency declaration.
The OROs referred to here are the State of Minnesota, Wright County, and Sherburne County. EAL SU6.3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
MNGP Basis Reference(s):
137
1. USAR Section 10.3.8, Plant Auxiliary Systems -Plant Service Systems, Plant Communications System 2. MNGP Emergency Plan Section 7 .2 -Communication Systems 3. MNGP Emergency Plan Figure 13.7-Direct Dedicated Telephones (Hot Lines) 138 APPENDIX A -ACRONYMS AND ABBREVIATIONS AC ......................................................................................................................
Alternating Current AOP .................................................................................................
Abnormal Operating Procedure APRM .................................................................................................
Average Power Range Meter ARI ...............................................................................................................
Alternate Rod Insertion ATWS ...................................................................................
Anticipated Transient Without Scram BWR .............................................................................................................
Boiling Water Reactor CDE ......................................................................................................
Com1nitted Dose Equivalent CFR ......................................................................................................
Code of Federal Regulations CTMT /CNMT ...............................................................................................................
Containment DBA ..............................................................................................................
Design Basis Accident DBE ........................................................................................................
... Design Basis Earthquake DC ..............................................................................................................................
Direct Current DSC .....................................................................................................................
Dry Shielded Cask EAL ...........................................................................................................
Emergency Action Level ECCS ............................................................................................
Emergency Core Cooling System ECL ................................................................................................
Emergency Classification Level EFT .............................................................................................................
Exhaust Filtration Train EOF ..................................................................................................
Emergency Operations Facility EOP ...............................................................................................
Emergency Operating Procedure EPA .............................................................................................
Environmental Protection Agency EPG ...............................................................................................
Emergency Procedure Guideline EPIP ................................................................................
Emergency Plan Implementing Procedure EPRI ................................
............................................................
Electric Power Research Institute ERG .................................................................................................
Emergency Response Guideline FAA ...............................................................................................
Federal Aviation Administration FBI ..................................................................................................
Federal Bureau of Investigation FEMA .............................................................................
Federal Emergency Management Agency FPB ..............................................................................................................
Fission Product Barrier FSAR ...................................................................................................
Final Safety Analysis Report GE ......................................................................................................................
General Emergency HCL ...................................................................................................................
Heat Capacity Limit HOO ................................................................................................
Headquarters Operation Officer HPCI ..............................................................................................
High Pressure Coolant Injection HSM .......................................................................................................
Horizontal Storage Module IC ........................................................................................................................
Initiating Condition ID .............................................................................................................................
Inside Diameter IPEEE. ............................
Individual Plant Examination of External Events (Generic Letter 88-20) ISFSI ...........................................................................
Independent Spent Fuel Storage Installation Keff ....................................................................................
Effective Neutron Multiplication Factor LCO ...............................................................................................
Limiting Condition of Operation LOCA ........................................................................................................
Loss of Coolant Accident MCR ..................................................................................................................
Main Control Room MSIV .....................................................................................................
Main Steam Isolation Valve MSL .......................................................................................................................
Main Steam Line mR, mRem, mrem, mREM ............................................................
milli-Roentgen Equivalent Man MW ....................................................................................................................................
Megawatt NEI .............................................................................................................
Nuclear Energy Institute A-1 NPP ..................................................................................................................
Nuclear Power Plant NRC ..............................................................................................
Nuclear Regulatory Commission NSSS .................................................................................................
Nuclear Steam Supply System NORAD .................................................................
North American Aerospace Defense Command NSPM ...................................................
Northern States Power Company, a Minnesota corporation
..........................................................................................................
doing business as Xcel Energy NUB .................................................................................................
Notification Of Unusual Event OBE .......................................................................................................
Operating Basis Earthquake OCA .............................................................................................................
Owner Controlled Area ODCM/ODAM
......................................................
Offsite Dose Calculation (Assessment)
Manual ORO ................................................................................................
Off-site Response Organization PA ..............................................................................................................................
Protected Area PAG .......................................................................................................
Protective Action Guideline PCIS ....................................................................................
Primary Containment Isolation System PRA/PSA ....................................
Probabilistic Risk Assessment I Probabilistic Safety Assessment PSIG .................................................................................................
Pounds per Square Inch Gauge R .........................................................................................................................................
Roentgen RCC ............................................................................................................
Reactor Control Console RCIC ...............................................................................................
Reactor Core Isolation Cooling RCS .............................................................................................................
Reactor Coolant System Rem, rem, REM ......................................................................................
Roentgen Equivalent Man RETS .......................................................................
Radiological Effluent Technical Specifications RHR .............................................................................................................
Residual Heat Removal RPS .........................................................................................................
Reactor Protection System RPV .............................................................................................................
Reactor Pressure Vessel RVLIS ........... ..........................................................
Reactor Vessel Level Instrumentation System RWCU ..........................................................................................
*..............
Reactor Water Cleanup SAE .................................................................................................................
Site Area Emergency SAMG .............................................................................
Severe Accident Management Guidelines SAR ..............................................................................................................
Safety Analysis Report SBO .........................................................................................................................
Station Blackout SCBA .....................................................................................
Self-Contained Breathing Apparatus SPDS ............................................................................................
Safety Parameter Display System SRO ............................................................................................................
Senior Reactor Operator SRV ....................................................................................................................
Safety Relief Valve TAF .....................................................................................................................
Top of Active Fuel TBNWS ...............................................................................
Turbine Building Normal Waste Sump . TED E .............................................................................................
Total Effective Dose Equivalent TSC ..........................................................................................................
Technical Support Center USAR .............................................................................................
Updated Safety Analysis Report A-2 APPENDIX B -DEFINITIONS The following definitions are taken from Title 10, Code of Federal Regulations, and related regulatory guidance documents.
Alert: Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels. General Emergency (GE): Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Notification of Unusual Event (NUE): Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs. Site Area Emergency (SAE): Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.
The following are key terms necessary for overall understanding the NEI 99-01 emergency classification scheme. Emergency Action Level (EAL): A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level. Emergency Classification Level (ECL): One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and off site response actions. The emergency classification levels, in ascending order of severity, are:
Notification of Unusual Event (NUE)
Alert
Site Area Emergency (SAE)
General Emergency (GE) Fission Product Barrier Threshold:
A pre-determined, site-specific, observable threshold B-1 indicating the loss or potential loss of a fission product barrier. Initiating Condition (IC): An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.
Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document.
The definitions of these terms are provided below. CONFINEMENT BOUNDARY:
The barrier(s) between areas containing spent fuel and the environment once the spent fuel is processed for dry storage. EXPLOSION:
A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.
Such events may require a post-event inspection to determine if the attributes of an explosion are present. FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.
IMMINENT:
The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. ISFSI PROTECTED AREA: The area surrounding the Independent Spent Fuel Storage Installation encompassed by the double chain link fence surrounding the ISFSI as defined in the Security Plan; the ISFSI Protected Area is excluded from the Plant Protected Area. B-2 NORMAL LEVELS: As applied to radiological IC/EALs, the highest reading in the past four hours excluding the current peak value. OWNER CONTROLLED AREA: The OCA boundaries consist of the plant property enclosed by a three strand barbed wire fence and a posted boundary on the Wright County side of the river. PROJECTILE:
An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA: The area surrounding the plant encompassed by the chain link fence and certain structures as defined in the Security Plan; excludes the ISFSI Protected Area. In areas where two fences are present, the inner fence is designated as the Protected Area barrier. REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool, or fuel transfer canal. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
SECURITY CONDITION:
Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION. SECONDARY CONTAINMENT:
SECONDARY CONTAINMENT includes the Reactor Building (including the HPCI Building), the Standby Gas Treatment System, the Offgas Dilution Fans, and connecting pipes and ducts. SECONDARY CONTAINMENT is isolated along with an automatic initiation of the Standby Gas Treatment System to minimize radiological releases to the environment.
SITE BOUNDARY:
For Dose Assessment and Protective Action Recommendation purposes the SITE BOUNDARY is the closest distance at which members of the public would be exposed to a radioactive release. The SITE BOUNDARY for liquid releases of radioactive material is defined in ODCM-02.01 (LIQUID EFFLUENTS).
The SITE BOUNDARY for gaseous releases of radioactive material is defined in ODCM-03.01 (GASEOUS EFFLUENTS).
UNISOLABLE:
An open or breached system line that cannot be isolated, remotely or locally. UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
B-3 L-MT-17-012 NSPM ATTACHMENT 5 MONTICELLO NUCLEAR GENERATING PLANT License Amendment Request to Revise the Emergency Action Level Scheme Emergency Action Level Matrix (For Information Only) (2 pages to follow)
The drawings specifically referenced in the table of contents have been processed into ADAMS. These drawings can be : accessed within the ADAMS : package or by performing a search'on the Document/Report Number. I : ! ' .; ------------
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