Text

License Amendment Request to Adopt Emergency Action Level Scheme Per NEI 99-01, Revision 6
	ML15044A245
Person / Time
Site:	Brunswick
Issue date:	01/30/2015
From:	Gideon W R; Duke Energy Progress
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15044A198	List: BSEP 15-0010, EAL-2 Modes 4, 5 & Defueled; ML15044A245; ML15044A246; ML15044A249;
References
	BSEP 15-0010
	Download: ML15044A245 (736)
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{{#Wiki_filter:William R. Gideon/lDUKE Vice President DUKE Brunswick Nuclear Plant ENERGY. P.O. Box 10429 Southport, NC 28461 0: 910.457.3698 January 30, 2015 10 CFR 50, Appendix E, IV.B.2 10 CFR 50.90 Serial: BSEP 15-0010 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 License Amendment Request to Adopt Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors" Ladies and Gentlemen: In accordance with the requirements of Part 50, Appendix E, Section IV.B.2 and 50.90 of Title 10 of the Code of Federal Regulations (CFR), Duke Energy Progress, Inc. (Duke Energy), is requesting approval of a proposed change in the emergency action levels (EALs) used at the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2.Duke Energy is proposing to change the emergency action levels from a scheme based on Revision 5 of NEI 99-01, "Methodology for Development of Emergency Action Levels," to one based on Revision 6 of NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors." Such a change in scheme requires NRC approval prior to implementation. The revised EALs will be implemented as two charts, one based on Operating Modes 1, 2, and 3 (i.e., hot operating conditions), and one based on Operating Modes 4, 5, and Defueled (i.e., cold operating conditions). Initiating conditions that are applicable to both hot and cold operating conditions are provided on both charts. This revision incorporates EALs for spent fuel instrumentation, which will be implemented after the instrumentation is installed, which will be Spring 2015 for Unit 2 and Spring 2016 for Unit 1.Enclosure 1 provides an evaluation of the proposed change to the EAL scheme. Enclosure 2 and 3 provide the Technical Bases Document, OPEP-02.2.1, "Emergency Action Level Technical Bases," both a redline and clean version respectively, for the proposed BSEP EALs.Enclosure 4 provides a comparison matrix between the NEI 99-01, Revision 6, EALs and the proposed BSEP EALs. This matrix identifies and provides justification for any differences from the NEI 99-01 EALs. Enclosure 5 provides the supporting calculation for the radiological effluent monitor EAL threshold values. Lastly, Enclosure 6 provides the proposed EAL wall charts, OPEP-02.1, "Initial Emergency Actions," for information. U.S. Nuclear Regulatory Commission Page 2 of 3 The proposed change contained in this submittal has been reviewed and approved by the Plant Nuclear Safety Committee (PNSC) and the amendment application has been approved for submission. Duke Energy requests approval for the proposed BSEP EALs by January 31, 2016, with a 180-day implementation period.In accordance with 10 CFR 50.91 (b)(1), Duke Energy is providing the State of North Carolina with a copy of the proposed license amendment. This document contains no regulatory commitments. Please refer any questions regarding this submittal to Mr. Kent Crocker, Manager -Emergency Preparedness, at (910) 457-3165.I certify under penalty of perjury that the foregoing is true and correct. Executed on January 30, 2015.Sincerely, William R. Gideon MKB/mkb

Enclosures:

1. Evaluation of the Proposed Change 2. BSEP Technical Bases Document, OPEP-02.2.1, "Emergency Action Level Technical Bases" (Redline Version)3. BSEP Technical Bases Document, OPEP-02.2.1, "Emergency Action Level Technical Bases" (Clean Version)4. BSEP EAL Comparison Matrix 5. Radiological Gaseous Effluent EAL Calculation, EP-EALCALC-BNP-0801, "Radiological Gaseous Effluent EAL Values Calculation (EALs RG1, RS1, RA1 and RU1)" 6. BSEP EAL Wall Charts, OPEP-02.1, "Initial Emergency Actions" U.S. Nuclear Regulatory Commission Page 3 of 3 cc (with enclosures):

U.S. Nuclear Regulatory Commission, Region II ATTN: Mr. Victor M. McCree, Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257 U.S. Nuclear Regulatory Commission ATTN: Mr. Andrew Hon (Mail Stop OWFN 8G9A)11555 Rockville Pike Rockville, MD 20852-2738 U.S. Nuclear Regulatory Commission ATTN: Ms. Michelle Catts, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 Mr. W. Lee Cox, Ill, Chief Radiation Protection Section NC Department of Health and Human Services 1645 Mail Service Center Raleigh, NC 27699-1645 Chair -North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 BSEP 15-0010 Enclosure 1 Page 1 of 7 Evaluation of the Proposed Change

Subject:

License Amendment Request to Adopt Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors" 1. Summary Description .................................................................................................... 2 2. Detailed Description ........................................................................................ .............. 2 2.1. Proposed Change .................................................................................................... 2 2 .2 .B ackg round .................................................................................................................. 2 3. Technical Evaluation .................................................................................................... 3 4. Regulatory Evaluation .................................................................................................. 3 4.1. Applicable Regulatory Requirements/Criteria .......................................................... 3 4 .2 .P re ce d e nt ..................................................................................................................... 4 4.3. No Significant Hazards Consideration Determination ............................................... 4 4 .4. C o nclusio ns .................................................................................................................. 6 5. Environmental Consideration ....................................................................................... 6 6. R efe re n ce s ......................................................................................................................... 7 BSEP 15-0010 Enclosure 1 Page 2 of 7 1. Summary Description In accordance with the provisions of Section 50, Appendix E, Section IV.B.2, and 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Duke Energy Progress, Inc. (Duke Energy), is proposing a change to the Brunswick Steam Electric Plant (BSEP) Emergency Plan by revising the Emergency Action Level (EAL) scheme.Duke Energy proposes to change the emergency action levels from a scheme based on Revision 5 of Nuclear Energy Institute (NEI) 99-01, "Methodology for Development of Emergency Action Levels," to a scheme based on Revision 6 of NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors." Such a change in scheme requires NRC approval prior to implementation. The proposed change would continue to meet the standards in 10 CFR 50.47(b) and the requirements in Appendix E to 10 CFR 50.2. Detailed Description

2.1. Proposed

Change The proposed change upgrades the BSEP EAL scheme to one based on Revision 6 of NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors." The EAL Technical Bases document (i.e., Enclosure 2 and 3) provides an explanation and rationale for the proposed change to each EAL.The BSEP Unit 1 and 2 NEI 99-01, Revision 6, EAL Comparison Matrix (i.e., Enclosure 4)provides a line-by-line comparison between the proposed BSEP Initiating Conditions, Mode Applicability, and EAL wording with the Initiating Conditions, Mode Applicability, and example EAL wording based on NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors." This document provides a means of assessing BSEP's differences and deviations from the NRC-endorsed guidance given in NEI 99-01, Revision 6.Discussion of BSEP EAL bases and lists of source document references are given in the EAL Technical Bases document (i.e., Enclosure 2 and 3). The EAL Technical Bases document provides background information for use with the BSEP Unit 1 and 2 NEI 99-01, Revision 6, EAL Comparison Matrix (i.e., Enclosure 4).2.2. Background EALs are the plant-specific indications, conditions, or instrument readings that are used to classify emergency conditions defined in the BSEP Emergency Plan.NEI 99-01, Revision 6, addresses lessons-learned since the implementation of NEI 99-01, Revision 5. In February 2008, NEI published NEI 99-01, Revision 5, in order to clarify the development guidance of numerous EALs, and enhance the guidance associated with the development of security-related EALs. In November 2012, NEI published NEI 99-01, Revision 6.The NRC formally endorsed the NEI 99-01, Revision 6, guidance as documented in a letter dated March 28, 2013 (i.e., Reference 1).NEI 99-01, Revision 6 (i.e., Reference 2), is the most recently accepted EAL methodology endorsed by the NRC. BSEP 15-0010 Enclosure 1 Page 3 of 7 BSEP's current EALs are based on NEI 99-01, Revision 5 guidance, as accepted by the NRC via letter dated November 6, 2009 (i.e., Reference 3).3. Technical Evaluation The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met. NEI 99-01 guidance methodology includes many years of development along with use and implementation. The guidance has been subject to NRC reviews and approval. The BSEP EAL scheme currently in place is based on the EAL methodology outlined in NEI 99-01, Revision 5. NEI 99-01, Revision 6 is the latest guidance endorsed by the NRC and provides guidance to nuclear power plant licensees for the development of a site-specific emergency classification scheme.10 CFR 50.47(b)(4) stipulates that Emergency Plans include a standard emergency classification and action level scheme. This scheme is a fundamental component of an Emergency Plan, in that it provides the defined thresholds that will allow site personnel to rapidly implement a range of pre-planned emergency response measures. An emergency classification scheme also facilitates timely decision-making by an offsite agency concerning the implementation of precautionary or protective actions for the public.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 NRC approval before implementing the change. Regulatory Issue Summary (RIS) 2005-02, Revision 1, "Clarifying the Process for Making Emergency Plan Changes," (i.e., Reference 4)also indicates that a revision to an entire EAL scheme, from NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," to another NRC-endorsed EAL scheme, must be submitted for prior NRC approval as specified in Section IV.B of Appendix E to 10 CFR 50.The proposed change to the EAL scheme to adopt the NEI 99-01, Revision 6, guidance, does not reduce the capability to meet the applicable emergency planning requirements established in 10 CFR 50.47 and 10 CFR 50, Appendix E.This change affects the BSEP Emergency Plan, but otherwise does not alter requirements of the Operating License or the Technical Specifications. This change does not alter any of the assumptions used in the safety analyses, nor does it cause any safety system parameters to exceed their acceptance limit. Therefore, the proposed change has no adverse effect on plant safety.Accordingly, pursuant to the requirements of 10 CFR 50, Appendix E, Section IV.B.2, Duke Energy requests NRC review and approval of the proposed change to the EAL scheme as provided in this license amendment request for BSEP, in accordance with 10 CFR 50.90.4. Regulatory Evaluation

4.1. Applicable

Regulatory Requirements/Criteria The regulations in 10 CFR 50.54(q) provide direction to licensees seeking to revise their Emergency Plan. The requirements related to nuclear power plant Emergency Plans are given BSEP 15-0010 Enclosure 1 Page 4 of 7 in the standards in 10 CFR 50.47, "Emergency Plans," and the requirements of Appendix E,"Emergency Planning and Preparedness for Production and Utilization Facilities," to 10 CFR 50.Paragraph (a)(1) of 10 CFR 50.47 states that no operating license for a nuclear power reactor will be issued unless a finding is made by the NRC that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. Section 50.47 establishes standards that onsite and offsite emergency response plans must meet for the NRC staff to make such a positive finding. One of these standards, 10 CFR 50.47(b)(4), stipulates that Emergency Plans include a standard emergency classification and action level scheme.Section IV.B to 10 CFR 50, Appendix E stipulates that Emergency Plans include EALs, which are to be used as criteria for determining the need for notification and participation of Local and State agencies, and for determining when and what type of protective measures should be considered to protect the health and safety of individuals both onsite and offsite. EALs are to be based on plant conditions and instrumentation, as well as onsite and offsite radiological monitoring. Section IV.B of Appendix E provides that initial EALs shall be discussed and agreed on by the applicant and State and local authorities. Section IV.B.2 to 10 CFR 50, Appendix E mandates a license amendment application is to be submitted for licensee's desiring to change its entire EAL scheme.NRC RIS 2005-02, Revision 1, issued in April 2011, also discusses that a change in an EAL scheme to incorporate the improvements provided in Nuclear Management and Resources Council, Inc./National Environmental Studies Project (NUMARC/NESP) 007, "Methodology for Development of Emergency Action Levels," or NEI 99-01 would not decrease the overall effectiveness of the Emergency Plan. However, due to the potential safety significance of the change, the change needs prior NRC review and approval. This approval would be granted via an NRC letter and supporting Safety Evaluation (SE).Duke Energy has determined that the proposed change does not require an exemption or relief from regulatory requirements and does not affect conformance with any General Design Criteria (GDC) differently than described in the BSEP Updated Final Safety Analysis Report (UFSAR).4.2. Precedent This request is similar in nature to requests from V.C. Summer Nuclear Station Unit 1 (ADAMS Accession No. ML14122A144); South Texas Project Units 1 and 2 (ADAMS Accession No.ML14164A341); Braidwood Stations, 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, Unit 1, Three Mile Island Nuclear Station, Unit 2 (ADAMS Accession No. ML14164A053), and Callaway Plant Unit 1 (ADAMS Accession No.ML14275A441). 4.3. No Significant Hazards Consideration Determination In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Duke Energy Progress, Inc. (Duke Energy), requests amendment to the Facility Operating License for Brunswick Steam Electric Plant (BSEP) to support the adoption of BSEP 15-0010 Enclosure 1 Page 5 of 7 Emergency Action Level (EAL) scheme based on Nuclear Energy Institute (NEI) 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," which has been endorsed by the NRC as documented in a letter dated March 28, 2013.The proposed change to the BSEP EAL scheme to adopt guidance in NEI 99-01, Revision 6 does not reduce the capability to meet the emergency planning requirements established in 10 CFR 50.47 and 10 CFR 50, Appendix E. The proposed change does not reduce the functionality, performance, or capability of BSEP's Emergency Response Organization (ERO) to respond in mitigating the consequences of accidents. The BSEP ERO functions will continue to be performed as required.The proposed change has been reviewed considering applicable requirements of 10 CFR 50.47, 10 CFR 50, Appendix E, and other applicable NRC documents. Duke Energy has evaluated the proposed change to the BSEP Emergency Plan and determined that the change does not involve a Significant Hazards Consideration. In support of this determination, an evaluation of each of the three standards, set forth in 10 CFR 50.92, is provided below.1) Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No The proposed change to the BSEP emergency action levels does not impact the physical function of plant structures, systems, or components (SSC) or the manner in which SCCs perform their design function. The proposed change does not authorize the addition of any new plant equipment or systems, nor does it alter the assumptions of any accident analyses. The proposed change does not adversely affect accident initiators or precursors, nor does it alter the design assumptions, conditions, and configuration or the manner in which the plant is operated and maintained. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No The proposed change to BSEP's EAL scheme to adopt the NRC-endorsed guidance in NEI 99-01, Revision 6, does not authorize any physical changes to the plant systems or equipment. The proposed change will not introduce failure modes that could result in a new accident, and the change does not alter assumptions made in the safety analysis.The proposed change will not alter the design configuration, or method of operation of plant equipment beyond its normal functional capabilities. The BSEP ERO functions will continue to be performed as required. The proposed change does not create any new credible failure mechanisms, malfunctions, or accident initiators. Therefore, the proposed change does not create the possibility of a new or different kind of accident from those that have been previously evaluated. BSEP 15-0010 Enclosure 1 Page 6 of 7 3) Does the proposed amendment involve a significant reduction in a margin of safety?Response: No The proposed change to BSEP's EAL scheme does not alter or exceed a design basis or safety limit. There is no change being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed change. The proposed change does not affect the Technical Specifications or the operating license. There are no changes to setpoints or environmental conditions of any SSC or the manner in which any SSC is operated. Margins of safety are unaffected by the proposed change to adopt the NEI 99-01, Revision 6, EAL scheme guidance. The applicable requirements or 10 CFR 50.47 and 10 CFR 50, Appendix E will continue to be met.Therefore, the proposed change does not involve any reduction in a margin of safety.4.4. Conclusions In conclusion, and based on the considerations discussed above, there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change to adopt the EAL scheme established in NEI 99-01, Revision 6, as endorsed by the NRC. Also, the change will be in compliance with the NRC regulations and 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. Environmental Consideration The proposed change is applicable to emergency planning requirements involving the proposed adoption of NRC-approved EAL guidance as described in NEI 99-01, Revision 6, and do not reduce the capability to meet the emergency planning standards established in 10 CFR 50.47 and 10 CFR 50, Appendix E. The proposed change does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in the individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change. BSEP 15-0010 Enclosure 1 Page 7 of 7 6. References

1) Letter from Mark Thaggard (U.S. Nuclear Regulatory Commission) to Susan Perkins-Grew (Nuclear Energy Institute), "U.S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 6," November 2012, dated March 18, 2013, ADAMS Accession Number ML12346A463

2) NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," dated November 2012, ADAMS Accession Number ML12326A805

3) Letter from Eric J. Leeds (U.S. Nuclear Regulatory Commission) to Benjamin Waldrep (Carolina Power & Light Company), "Brunswick Steam Electric Plant, Units 1 and 2 -Revision to Emergency Action Levels (TAC Nos. ME01 17 and ME01 18)," dated November 6, 2009, ADAMS Accession Number ML092680102

4) Regulatory Issue Summary 2005-02, Revision 1, "Clarifying the Process for Making Emergency Plan Changes," dated August 19, 2011, ADAMS Accession Number ML100340545 BSEP 15-0010 Enclosure 2 Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 License Amendment Request to Adopt Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors" BSEP Technical Bases Document, OPEP-02.2.1,"EmerQencv Action Level Technical Bases" (Redline Version)

DUKE, R ENERGY. BRUNSWICK NUCLEAR PLANT Reference Use PLANT OPERATING MANUAL VOLUME XIII PLANT EMERGENCY PROCEDURE OPEP-02.2.1 EMERGENCY ACTION LEVEL TECHNICAL BASES REVISION 6 (Draft D4 8/16/14)I OPEP-02.2.1 Rev. 6 1 Page 1 of 3101 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE ................................................................................................................................... 3 2.0 DISCUSSION .............................................................................................................................. 3 2.1 Background .............................................................................................................................. 3 2.2 Fission Product Barriers ........................................................................................................ 4 2.3 Fission Product Barrier Classification Criteria ...................................................................... 4 2.4 EAL Organization ..................................................................................................................... 5 2.5 Technical Bases Inform ation ................................................................................................. 7 2.6 Operating Mode Applicability ............................................................................................... 8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ............................................... 9 3.1 General Considerations ........................................................................................................ 9 3.2 Classification Methodology ................................................................................................. 10

4.0 REFERENCES

.......................................................................................................................... 14 4.1 Developm ental ....................................................................................................................... 14 4.2 Im plementing .............................................................................................................................. 5.0 DEINITIONS, ACRONYM S & ABBREVIATIO NS ................................................................. 15 6.0 BNP TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE ......................................................... 22 7.0 ATTACHM ENTS ....................................................................................................................... 26 1 Em ergency Action Level Technical Bases ............................................................... 27 Category R Abnorm al Rad Release / Rad Effluent ......................................... 27 Category C Cold Shutdown / Refueling System Malfunction ........................... 74 Category H Hazards .......................................................................................... 128 Category S System M alfunction ........................................................................ 182 Category E ISFSI ............................................................................................... 232 Category F Fission Product Barrier Degradation ............................................... 236 2 Fission Product Barrier Loss / Potential Loss Matrix and Bases ........................................................................................................ 243 3 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases ..................................... 306 I OPEP-02.2.1 I Rev. 6 1 Page 2 of 310J

1.0 PURPOSE

This document provides an explanation and rationale for each Emergency Action Level (EAL)included in the EAL Upgrade Project for Brunswick Nuclear Plant (BNP). It should be used to facilitate review of the BNP EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of 0PEP-02.1 Initial Emergency Actions, may use this document as a technical reference in support of EAL interpretation. This information may assist the Site Emergency Coordinator in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to offsite officials. The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.

2.0 DISCUSSION

2.1 Background

EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the BNP Emergency Plan.In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance.NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included: " Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.

Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSIs)." Simplifying the fission product barrier EAL threshold for a Site Area Emergency.

Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs).Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ADAMS Accession Number ML1 10240324) (ref.4.1.1), BNP conducted an EAL implementation upgrade project that produced the EALs discussed herein 0PEP-02.2.1 Rev. 6 Page 3 of 310

2.2 Fission

Product Barriers 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. Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. "Loss" and"Potential Loss" signify the relative damage and threat of damage to the barrier. A "Loss" threshold means the barrier no longer assures containment of radioactive materials. A"Potential Loss" threshold implies an increased probability of barrier loss and decreased certainty of maintaining the barrier.The primary fission product barriers are: A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.B. Reactor Coolant System (RCS): 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.C. Primary Containment (PC): 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. Primary Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.

2.3 Fission

Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss: Alert: Any loss or any potential loss of either Fuel Clad or RCS barrier Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of the third barrier OPEP-02.2.1 ev. 6 Page 4 of 310 2.4 EAL Organization The BNP EAL scheme includes the following features: " Division of the EAL set into three broad groups: o EALs applicable under all plant operating modes -This group would be reviewed by the EAL-user any time emergency classification is considered. o EALs applicable only under hot operating modes -This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup, or Power Operation mode.o EALs applicable only under cold operating modes -This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode.The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency." Within each group, assignment of EALs to categories and subcategories: Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The BNP EAL categories are aligned to and represent the NEI 99-01"Recognition Categories." Subcategories are used in the BNP scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The BNP EAL categories and subcategories are listed below.OPEP-02.2.1 Rev. 6 Page 5 of 310 EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory Any Operating Mode: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent 2 -Irradiated Fuel Event 3 -Area Radiation Levels H -Hazards and Other Conditions 1 -Security Affecting Plant Safety 2 -Seismic Event 3 -Natural or Technological Hazard 4 -Fire 5 -Hazardous Gas 6 -Control Room Evacuation 7 -SEC Judgment E -Independent Spent Fuel Storage 1 -Confinement Boundary Installation (ISFSI)Hot Conditions: S -System Malfunction 1 -Loss of Emergency AC Power 2 -Loss of Vital DC Power 3 -Loss of Control Room Indications 4- RCS Activity 5- RCS Leakage 6 -RPS Failure 7 -Loss of Communications 8 -Hazardous Event Affecting Safety Systems F -Fission Product Barrier Degradation None Cold Conditions: C -Cold Shutdown / Refueling System 1 -RPV Level Malfunction 2 -Loss of Emergency AC Power 3 -RCS Temperature 4 -Loss of Vital DC Power 5 -Loss of Communications 6 -Hazardous Event Affecting Safety Systems The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL Technical Bases Document in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachments 1 & 2 of this document for such information. OPEP-02.2.1 Rev. 6 Page 6 of 310

2.5 Technical

Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (R, C, H, S, F and E) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided: Category Letter & Title Subcategory Number & Title Initiating Condition (IC)Site-specific description of the generic IC given in NEI 99-01 Rev. 6.EAL Identifier (enclosed in rectanale) Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:

1. First character (letter):

Corresponds to the EAL category as described above (R, C, H, S, F or E)2. Second character (letter): The emergency classification (G, S, A or U)G = General Emergency S = Site Area Emergency A = Alert U = Unusual Event 3. Third character (number): Subcategory number within the given category.Subcategories are sequentially numbered beginning with the number one (1). If a category does not have a subcategory, this character is assigned the number one (1).4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1).Classification (enclosed in rectangle): Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G)EAL (enclosed in rectangle) Exact wording of the EAL as it appears in the EAL Classification Matrix Mode Applicability OPEP-02.2.1 Rev. 6 Page 7 of 310 One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown, 4 -Cold Shutdown, 5 -Refueling, D -Defueled, or All. (See Section 2.6 for operating mode definitions) Definitions: If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1.Basis: A Plant-Specific basis section that provides BNP-relevant information concerning the EAL.This is followed by a Generic basis section that provides a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.BNP Basis Reference(s): Site-specific source documentation from which the EAL is derived 2.6 Operating Mode Applicability (ref. 4.1.7)1 Power Operations Reactor is critical and the mode switch is in RUN 2 Startup The mode switch is in STARTUP/HOT STANDBY 3 Hot Shutdown The mode switch is in SHUTDOWN, all reactor vessel head closure bolts are fully tensioned, and reactor coolant temperature is >212'F 4 Cold Shutdown The mode switch is in SHUTDOWN, all reactor vessel head closure bolts are fully tensioned, and reactor coolant temperature is < 212°F 5 Refuel The mode switch is in REFUEL or SHUTDOWN and one or more reactor vessel head closure bolts are less than fully tensioned D Defueled RPV contains no irradiated fuel The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.IPEP-02.2.1 Rev. 6 Page 8 of 310

3.0 GUIDANCE

ON MAKING EMERGENCY CLASSIFICATIONS

3.1 General

Considerations When making an emergency classification, the Site Emergency Coordinator 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 based on loss or potential loss of Fission Product Barrier Thresholds.

3.1.1 Classification

Timeliness 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" (ref. 4.1.12).3.1.2 Valid Indications All emergency classification assessments shall 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, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

3.1.3 Imminent

Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Site Emergency Coordinator 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.3.1.4 Planned vs. Unplanned Events 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 OPEP-02.2.1 IRev. 6 1Page 9 of 310 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 (ref.4.1.4).3.1.5 Classification Based on Analysis 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.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift). 3.1.6 Site Emergency Coordinator Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Site Emergency Coordinator 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 Site Emergency Coordinator 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 in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.3.2 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 must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process "clock" starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process "clock" started.When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock." For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 4.1.14).3.2.1 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: OPEP-02.2.1 Rev. 6 Page 10 of 310

If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two different units, a Site Area Emergency should be declared.There is no "additive" effect from multiple EALs meeting the same ECL. For example:* If two Alert EALs are met, whether at one unit or at two different units, an Alert should be declared.Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.2).3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable.

If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode)is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher.3.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Site Emergency Coordinator 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 Site Emergency Coordinator, 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.3.2.4 Emergency Classification Level Upgrading and Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s)and EAL(s). The ECL may also simply be terminated. As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.2).3.2.5 Classification of Short-Lived Events 0PEP-02.2.1 I Rev. 6 Page 11 of 310 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 the reactor followed by a successful manual scram.3.2.6 Classification of Transient Conditions Many of the ICs and/or EALs 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 durinq expected plant response -In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures. EAL momentarily met but the condition is corrected prior to an emergency declaration -If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example: An ATWS occurs and the high pressure ECCS systems fail to automatically start. RPV level rapidly decreases and the plant enters an inadequate core cooling condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be based on the ATWS only.It is important to stress that the 15-minute emergency classification assessment period (process clock) 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 when an operator is able to take a successful corrective action prior to the Site Emergency Coordinator 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. 3.2.7 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 OPEP-02.2.1 Rev. 6 1 Page 12 of 310 condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 (ref. 4.1.4) 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.

3.2.8 Retraction

of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3).IoPEP-02.2.1 I Rev. 6 Page 13 of 310

4.0 REFERENCES

4.1 Developmental

4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML1 10240324 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007.4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 § CFR 50.73 License Event Report System 4.1.6 BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan 4.1.7 BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations 4.1.8 Technical Specifications Table 1.1-1 Modes 4.1.9 Technical Specifications Section 3.6 Containment Systems 4.1.10 PRO-NGGC-0201 NGG Procedure Writers Guide 4.1.11 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.12 NGGM-PM-0028 Transnuclear NUHOMS Dry Fuel Storage Program Manual 4.2 Implementing 4.2.1 OPEP-02.1 Initial Emergency Actions 4.2.2 NEI 99-01 Rev. 6 to BNP EAL Comparison Matrix 4.2.3 BNP EAL Matrix IOPEP-02.2.1 I Rev. 6 1 Page 14 of 310 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS

5.1 Definitions

(ref. 4.1.1 except as noted)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.Can/Cannot Be Maintained Above/Below The value of an identified parameter is/is not able to be held within the specified limit. The determination requires an evaluation of system performance and availability in relation to parameter values and trends. An instruction prescribing action when a parameter cannot be maintained above or below a specified limit neither requires nor prohibits anticipatory action-depending upon plant conditions, the action may be taken as soon as it is determined that the limit will ultimately be exceeded, or delayed until the limit is actually reached. Once the parameter does exceed the limit, however, the action must be performed; it may not be delayed while attempts are made to restore the parameter to within the desired control band.Can/Cannot Be Restored Above/Below The value of an identified parameter is/is not able to be brought within the specified limit. The determination requires an evaluation of system performance and availability in relation to parameter values and trends. An instruction prescribing action when a value cannot be restored and maintained above or below a specified limit does not require immediate action simply because the current values is outside the range, but does not permit extended operation beyond the limit; the action must be taken as soon as it is apparent that the specified range cannot be attained.Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the BNP ISFSI, Confinement Boundary is defined as the Dry Shielded Canister (DSC) (Ref. 4.1.12).Containment Closure The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to BNP, Containment Closure is established when either Primary Containment is Operable per Section 3.6.1.1 of Technical Specifications or Secondary Containment is considered functional per the requirements of OAP-022, BNP Outage Risk Management. EPA PAGs Environment Protection Agency Protective Action Guidelines. The EPA PAGs are expressed in terms of dose commitment: 1 Rem TEDE or 5 Rem CDE Thyroid. Actual or projected offsite OPEP-02.2.1 Rev. 6 Page 15 of 310 exposures in excess of the EPA PAGs requires BNP to recommend protective actions for the general public to offsite planning agencies.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 require a post-event inspection to determine if the attributes of an explosion are present.Deleted: Faultedjl fire ------ -The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or generator pressure or the steam generator to overheated electrical equipment do not constitute fires. Observation of flame is preferred but become completely depressurized.¶ is NOT required if large quantities of smoke and heat are observed.Flooding A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.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 BNP 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 BNP. 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.Impede(d)OPEP-02.2.1 I Rev. 6 1 Page 16 of 310 Personnel access to a room or area is hindered to an extent that 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). Intrusion The act of entering without authorization. Discovery of a bomb in a specified area is indication of intrusion into that area by a hostile force.Maintain Take appropriate action to hold the value of an identified parameter within specified limits.Normal Levels As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value.Owner Controlled Area Area depicted as the property boundary in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan (ref. 4.1.6).Projectile An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.Protected Area The double-fenced security area with intrusion detection devices immediately surrounding the plant structures. The Protected Area is depicted in BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations (ref.4.1.7).RCS Intact The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).Refueling Pathway The reactor refueling cavity, spent fuel pool and fuel transfer canal (cattle chute) comprise the refueling pathway.Restore Take the appropriate action required to return the value of an identified parameter to the applicable limits Safety System OPEP-02.2.1 Rev. 6 Page 17 of 310 , I Deleted: Ruptured[] _ The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.¶ 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 (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. 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.Site Boundary Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan (ref. 4.1.6).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.Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. 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. I OPEP-02.2.1 Rev. 6 1 Page 18 of 310 5.2 Abbreviations/Acronyms 0 F ....................................................................................................... Degrees Fahrenheit 0 ........................................................................................................................... Degrees AC ....................................................................................................... Alternating Current AO P ................................................................................. Abnorm al Operating Procedure APRM .................................................................................. Average Power Range M eter ATW S ..................................................................... Anticipated Transient W ithout Scram BNP ............................................................................................. Brunswick Nuclear Plant BW R .............................................................................................. Boiling W ater Reactor BW ROG ................................................................. Boiling W ater Reactor Owners G roup CDE ...................................................................................... Com m itted Dose Equivalent CFR ...................................................................................... Code of Federal Regulations CS ................................................................................................................... Core Spray DBA ................................................................................................ Design Basis Accident DC ............................................................................................................... Direct Current EAL ............................................................................................. Em ergency Action Level ECCS ............................................................................ Em ergency Core Cooling System ECL .................................................................................. Em ergency Classification Level EO F ................................................................................... Em ergency Operations Facility EO P .............................................................................. Em ergency Operating Procedure EPA ............................................................................... Environm ental Protection Agency EPG ............................................................................... Em ergency Procedure G uideline EPIP ................................................................ Em ergency Plan Im plem enting Procedure ESF ......................................................................................... Engineered Safety Feature FAA .................................................................................. Federal Aviation Adm inistration FBI ................................................................................... Federal Bureau of Investigation FEMA .............................................................. Federal Em ergency Managem ent Agency FSAR ................................................................................... Final Safety Analysis Report G E ...................................................................................................... General Em ergency HCTL ............................................................................. Heat Capacity Tem perature Lim it HPCI ............................................................................... High Pressure Coolant Injection IC ......................................................................................................... Initiating Condition IPEEE ................ Individual Plant Examination of External Events (Generic Letter 88-20)ISFSI ........................................................... Independent Spent Fuel Storage Installation OPEP-02.2.1 I Rev. 6 1 Page 19 of 310 Keff ......................................................................... Effective Neutron M ultiplication Factor LCO .................................................................................. Lim iting Condition of O peration LER ............................................................................................... Licensee Event Report LOCA ......................................................................................... Loss of Coolant Accident LPSI ................................................................................... Low Pressure Safety Injection LW R .................................................................................................. Light W ater Reactor MPC ................................... Maximum Permissible Concentration/Multi-Purpose Canister M PH ........................................................................................................... M iles Per Hour M SIV ...................................................................................... M ain Steam Isolation Valve M SL ......................................................................................................... M ain Steam Line m R, m Rem , m rem , m REM .............................................. m illi-Roentgen Equivalent M an M W .................................................................................................................... M egawatt NEI .............................................................................................. Nuclear Energy Institute NESP .................................................................. National Environm ental Studies Project NPP ................................................................................................... Nuclear Power Plant NRC ............................................................................... Nuclear Regulatory Com m ission NSSS ................................................................................ Nuclear Steam Supply System NO RAD .................................................. North Am erican Aerospace Defense Com m and (NO )UE ............................................................................... Notification of Unusual Event O BE ...................................................................................... O perating Basis Earthquake O CA .............................................................................................. Owner Controlled Area ODCM/ODAM ......................................... Offsite Dose Calculation (Assessment) Manual O RO .................................................................................. Offsite Response O rganization PA .............................................................................................................. Protected Area PRA/PSA ..................... Probabilistic Risk Assessment / Probabilistic Safety Assessment PW R ...................................................................................... Pressurized W ater Reactor PSIG ............................................................................... Pounds per Square Inch G auge R ........................................................................................................................ Roentgen RB ........................................................................................................... Reactor Building RCIC ................................................................................. Reactor Core Isolation Cooling RCS ............................................................................................ Reactor Coolant System Rem , rem , REM ...................................................................... Roentgen Equivalent M an RETS ........................................................ Radiological Effluent Technical Specifications RPS ......................................................................................... Reactor Protection System OPEP-02.2.1 Rev. 6 1 Page 20 of 310 RPV ............................................................................................ Reactor Pressure Vessel RW CU .......................................................................................... Reactor W ater Cleanup SAR ............................................................................................... Safety Analysis Report SBGTS .......................................................................... Stand-By Gas Treatment System SBO ......................................................................................................... Station Blackout SCBA ...................................................................... Self-Contained Breathing Apparatus SEC ...................................................................................... Site Emergency Coordinator SPDS ........................................................................... Safety Parameter Display System SRO ........................................................................................... Senior Reactor Operator TEDE ............................................................................... Total Effective Dose Equivalent TAF ....................................................................................................... Top of Active Fuel TSC ........................................................................................... Technical Support Center OPEP-02.2.1 6 Page 21 of 310 6.0 BNP-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a BNP EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the BNP EALs based on the NEI guidance can be found in the EAL Comparison Matrix.BNP NEI 99-01 Rev. 6 EAL IC Example EAL RU1.1 AU1 1,2 RU1.2 AU1 3 RU2.1 AU2 1 RA1.1 AA1 1 RA1.2 AA1 2 RA1.3 AA1 3 RA1.4 AA1 4 RA2.1 AA2 1 RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1 RA3.2 AA3 2 RS1.1 AS1 1 RS1.2 AS1 2 RS1.3 AS1 3 RS2.1 AS2 1 RGI.1 AG1 1 RG1.2 AG1 2 RG1.3 AGI 3 RG2.1 AG2 1 OPEP-02.2.1 Rev. 6 Page 22 of 310 BNP NEI 99-01 Rev. 6 EAL IC Example EAL CUl.1 cui 1 CU1.2 Cui 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1,2,3 CA1.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1,2 CA6.1 CA6 1 CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 1 CG1.2 CG1 2 FA1.1 FA1 1 FS1.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1 HU1.2 HU1 2 HU1.3 HU1 3 HU2.1 HU2 1 HU3.1 HU3 1 IOPEP-02.2.1 I Rev. 6 Page 23 of 310 BNP NEI 99-01 Rev. 6 EAL IC Example EAL HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 HU3.5 HU3 5 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HAl 1 HA1.2 HA1 2 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA7 1 HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 HG1.1 HG1 1 HG7.1 HG7 1 SuI.1 Sul 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1,2,3 SU6.1 SU5 1 I OPEP-02.2.1 Rev. 6 1 Page 24 of 310 BNP NEI 99-01 Rev. 6 EAL IC Example EAL SU6.2 SU5 2 SU7.1 SU6 1,2,3 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA8.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SG8 1 EU1.1 E-HU1 1 IOPEP-02.2.1 I Rev. 6 1 Page 25 of 310

7.0 ATTACHMENTS

7.1 Attachment

1, Emergency Action Level Technical Bases 7.2 Attachment 2, Fission Product Barrier Matrix and Basis I OPEP-02.2.1 I Rev. 6 1 Page 26 of 310 ATTACHMENT 1 Page 1 of 219 EAL Bases Category R -Abnormal Rad Release / Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification. At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.Events of this category pertain to the following subcategories:

1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.

3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.

1OPEP-02.2.1 I Rev. 6 Page 27 of 310 ATTACHMENT 1 Page 2 of 219 EAL Bases Category: R -Abnormal Rad Levels / Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity > 2 times the ODCM limits for 60 minutes or longer EAL: RU1.1 Unusual Event Reading on any Table R-1 effluent radiation monitor > column "UE" for -> 60 min.(Notes 1, 2, 3)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE I SAE [ Alert UE Main Stack Red D12-RM-23S 2.13E+09 pCi/sec 2.13E+08 pCi/sec 2.13E+07 pCi/sec 1.80E+06 p.Ci/sec Reactor Bldg Vent Noble CAC-AQH-1264-3

........

6.14E+04 cpm o Gas Turbine Bldg Vent D12-RM-23 1.07E+08 pCi/sec 1.07E+07 pCi/sec 1.07E+06 pCi/sec 1.13E+04 pCi/sec Service Water Effluent D12-RM-K605 -- .....---- 2 x hi alarm Rad R-i Radwaste Effluent Rad D12-RM-K604

.........

2 x hi-hi alarm Mode Applicability: All I OPEP-02.2.1 I Rev. 6 1 Page 28 of 310 ATTACHMENT 1 Page 3 of 219 EAL Bases Definition(s): None BNP Basis: Gaseous Releases The column "UE" gaseous release values in Table R-1 represent two times the appropriate ODCM release rate limits associated with the specified monitors (ref. 2, 3).Instrumentation that may be used to assess this EAL is listed below (ref. 1):* Main Stack Monitoring System Noble Gas Activity Monitor 2-D1 2-RM-23S (1/2-D12-RR-4599-4)" Reactor Building Ventilation Monitoring System Noble Gas Activity Monitor 1(2)-CAC-AQH-1264-3

Turbine Building Ventilation Monitoring System Noble Gas Activity Monitor 1(2)-D1 2-RM-23 (1/2-DL12-RR-4548-4)

Liquid Releases Instrumentation that may be used to assess this EAL is listed below:* Liquid Radwaste Radioactivity Monitor 2-D12-RM-K604 (batch release)" Main Service Water Effluent Radioactivity Monitor 1(2)-D12-RM-K605 (continuous release)The Liquid Radwaste Radioactivity Monitor Hi-Hi alarm automatically closes Radwaste Liquid Effluent Discharge Valves D12-V27A and 27B. The Hi-Hi alarm setpoint is set in accordance with the ODCM and includes a conservative reduction factor of 20 to the ODCM release rate limit (ref. 1, 2).The Main Service Water Effluent Radioactivity Monitor High alarm setpoint is set in accordance with the ODCM and ensures continuous liquid releases do not exceed ODCM Section 7.3.3 limits.O 0PEP-02.2.1 I Rev. 6 1 Page 29 of 310 1 ATTACHMENT 1 Page 4 of 219 EAL Bases NEI 99-01 Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.Nuclear 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.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 Deleted: EAL #1 -Deleted: EAL #2 -This EAL addresses release limits for 30 minutes does not meet the EAL. radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge ,This EAL addresses normally occurring continuous radioactivity releases from monitored permit. This EAL will typically be associated -with planned batch releases from non-lncontinuous release pathways (e.g., radwaste, gaseous or liquid effluent pathways. waste gas).¶I EAL #3 -This EAL addresses uncontrolled gaseous or liquid releases that are detected by , llation of the emergency classification level would be via IC.RAl ..-----------. sample analyses or environmental surveys, sli particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, BNP Basis Reference(s): t etc.).¶Deleted: AA1 I oPEP-02.2.1 Rev. 6 1 Page 30 of 310

1. BNP ODCM Appendix E Radioactive Liquid and Gaseous Effluent Monitoring Instrumentation Numbers 2. BNP Offsite Dose Calculation Manual 3. EP-EALCALC-BNP-0801 Radiological Gaseous Effluent Values (EALs RG1, RS1, RA1 and RU1)4. NEI 99-01 AU1 I OPEP-02.2.1 I Rev. 6 1 Page 31 of 310 ATTACHMENT 1 Page 6 of 219 EAL Bases Category:

R -Abnormal Rad Levels / Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.EAL: RU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate> 2 x ODCM limits for > 60 min. (Notes 1, 2)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Mode Applicability: All Definition(s): None BNP Basis: None NEI 99-01 Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.Nuclear 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.OPEP-02.2.1 IRev. 6 Page 32 of 310 ATTACHMENT 1 Page 7 of 219 EAL Bases 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. IReleases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.jThis -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 ICRA. ............... BNP Basis Reference(s):

1. BNP ODCM Appendix E Radioactive Liquid and Gaseous Effluent Monitoring Instrumentation Numbers 2. BNP Offsite Dose Calculation Manual 3. NEI 99-01 AU1 Deleted: 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.¶ Deleted: EAL #1 -This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.¶ EAL #2 -This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).¶EAL #3 -Deleted: AA1 I OPEP-02.2.1 I Rev. 6 1 Page 33 of 310 ATTACHMENT 1 Page 8 of 219 EAL Bases Category: Subcategory: Initiating Condition: EAL: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE RA1.1 Alert In the absence of real-time dose assessment, reading on any Table R-1 effluent radiation monitor > column "ALERT" for a 15 min. (Notes 1, 2, 3, 4)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4 The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Main Stack Rad D12-RM-23S 2.13E+09 pCi/sec 2.13E+08 pCi/sec 2.13E+07 pCi/sec 1.80E+06 pCi/sec Reactor Bldg Vent Noble CAC-AQH-1264-3 ............ 6.14E+04 cpm Gas Turbine Bldg Vent D12-RM-23 1.07E+08 pCi/sec 1.07E+07 pCi/sec 1.07E+06 pCi/sec 1.13E+04 pCi/sec Service Water Effluent D12-RM-K605

---- .... 2 x hi alarm Rad R7 Radwaste Effluent Rad D12-RM-K604

.... .... 2 x hi-hi alarm Mode Applicability:

All I OPEP-02.2.1 I Rev. 6 Page 34 of 310 ATTACHMENT 1 Page 9 of 219 EAL Bases Definition(s): None BNP Basis: This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either (ref. 2, 3):* 1OmRemTEDE

50 mRem CDE Thyroid The column "ALERT" gaseous effluent release values in Table R-1 correspond to calculated doses of 1% (10% of the SAE thresholds) of the EPA Protective Action Guidelines (TEDE or CDE Thyroid).Instrumentation that may be used to assess this EAL is listed below (ref 1): " Main Stack Monitoring System Noble Gas Activity Monitor 2-D12-RM-23S (1/2-D12-RR-4599-4)* Turbine Building Ventilation Monitoring System Noble Gas Activity Monitor 1(2)-D1 2-RM-23 (1/2-DL12-RR-4548-4)

NEI 99-01 Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.OPEP-02.2.1 Rev. 6 Page 35 of 310 ATTACHMENT 1 Page 10 of 219 EAL Bases 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 ICS. ................ BNP Basis Reference(s):

1. BNP ODCM Appendix E Radioactive Liquid and Gaseous Effluent Monitoring Instrumentation Numbers 2. BNP Offsite Dose Calculation Manual 3. EP-EALCALC-BNP-0801 Radiological Gaseous Effluent Values (EALs RG1, RS1, RA1 and RU1)4. NEI 99-01 AA1 , j Deleted: ASI IOPEP-02.2.1 IRev. 6 1Page 36 of 310 ATTACHMENT 1 Page 11 of 219 EAL Bases R -Abnormal Rad Levels / Rad Effluent Category: Subcategory:

1 -Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: RA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 3, 4)Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All Definition(s): SITE BOUNDARY-Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan BNP Basis: Dose assessments are performed by computer-based methods (ref. 1, 2, 3)IOPEP-02.2.1 I Rev. 6 Page 37 of 310 ATTACHMENT 1 Page 12 of 219 EAL Bases NEI 99-01 Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Escalation of the emergency classification level would be via ICB.S1 ---Deleted:---------- BNP Basis Reference(s):

1. EMG-NGGC-0002 Off-site Dose Assessment

2. 0PEP-03.4.7 Automation of Off-Site Dose Projections

3. OE&RC-03.4.8, Offsite Dose Projections for Monitored Releases 3. NEI 99-01 AA1 IPEP-02.2.1 ev. 6 Page 38 of 310 ATTACHMENT 1 Page 13 of 219 EAL Bases R -Abnormal Rad Levels / Rad Effluent Category: Subcategory:

1 -Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: RA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Mode Applicability: All Definition(s): SITE BOUNDARY-Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan BNP Basis: Dose assessments based on liquid releases are performed per Offsite Dose Calculation Manual (ref. 1).IoPEP-02.2.1 I Rev. 6 Page 39 of 310 ATTACHMENT 1 Page 14 of 219 EAL Bases NEI 99-01 Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Escalation of the emergency classification level would be via ICRS1 .................. BNP Basis Reference(s):

1. BNP Offsite Dose Calculation Manual 2. NEI 99-01 AA1 ,- Deleted: AS1 OPEP-02.2.1 Rev. 6 Page 40 of 310 ATTACHMENT 1 Page 15 of 219 EAL Bases Category: Subcategory:

Initiating Condition: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: RA1.4 Alert Field survey results indicate EITHER of the followingat orbeyond the SITE BOUNDARY:* Closed window dose rates > 10 mR/hr expected to continue for > 60 min.* Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation.(Notes 1,2)I -i Formatted: Font: Bold, No underline Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Mode Applicability: All Definition(s): SITE BOUNDARY-Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan BNP Basis: OPEP-02.6.6, Environmental Monitoring Team Leader and 0PEP-03.5.5 Environmental Monitoring and Plume Tracking provide guidance for emergency or post-accident radiological environmental monitoring (ref. 1, 2).IOPEP-02.2.1 I Rev. 6 1 Page 41 of 310 ATTACHMENT 1 Page 16 of 219 EAL Bases NEI 99-01 Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Escalation of the emergency classification level would be via ICBS1.................. BNP Basis Reference(s):

1. OPEP-02.6.6, Environmental Monitoring Team Leader 2. 0PEP-03.5.5 Environmental Monitoring and Plume Tracking 3. NEI 99-01 AA1--Deleted: AS1 I OPEP-02.2.1 Rev. 6 Page 42 of 310 ATTACHMENT 1 Page 17 of 219 EAL Bases Category:

R -Abnormal Rad Levels / Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL: RS1.1 Site Area Emergency In the absence of real-time dose assessment, reading on any Table R-1 effluent radiation monitor > column "SAE" for > 15 min.(Notes 1,2, 3, 4)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAi.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE I Alert UE Main Stack Rad D12-RM-23S 2.13E+09 pCi/sec 2.13E+08 pCi/sec 2.13E+07 pCi/sec 1.80E+06 IpCi/sec Reactor Bldg Vent Noble CAC-AQH-1264-3 --- .... ..... 6.14E+04 cpm Gas Turbine Bldg Vent D12-RM-23 1.07E+08 pCi/sec 1.07E+07 pCi/sec 1.07E+06 pCi/sec 1.13E+04 pCi/sec Service Water Effluent D12-RM-K605 2 x hi alarm Rad Radwaste Effluent Rad D12-RM-K604

.... ---- 2 x hi-hi alarm Mode Applicability:

All OPEP-02.2.1 Rev. 6 Page 43 of 310 ATTACHMENT 1 Page 18 of 219 EAL Bases Definition(s): None BNP Basis: This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either (ref. 2):* 100mRemTEDE

500 mRem CDE Thyroid The column "SAE" gaseous effluent release values in Table R-1 correspond to calculated doses of 10% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid).Instrumentation that may be used to assess this EAL is listed below (ref 1):* Main Stack Monitoring System Noble Gas Activity Monitor 2-D12-RM-23S (1/2-D12-RR-4599-4)" Turbine Building Ventilation Monitoring System Noble Gas Activity Monitor 1(2)-D12-RM-23 (1/2-DL1 2-RR-4548-4)

NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that 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 em ergency classification level would be via IC RG 1. ..................--- -Deleted: AGI OPEP-02.2.1 I Rev. 6 1 Page 44 of 310 ATTACHMENT 1 Page 19 of 219 EAL Bases BNP Basis Reference(s):

1. BNP ODCM Appendix E Radioactive Liquid and Gaseous Effluent Monitoring Instrumentation Numbers 2. EP-EALCALC-BNP-0801 Radiological Gaseous Effluent Values (EALs RG1, RS1, RA1 and RU1)3. NEI 99-01 AS1 iOPEP-02.2.1 I Rev. 1 Page 45 of 310 ATTACHMENT 1 Page 20 of 219 EAL Bases Category:

R -Abnormal Rad Levels / Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL: RS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 3, 4)Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All Definition(s): SITE BOUNDARY-Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan BNP Basis: Dose assessments are performed by computer-based methods (ref. 1, 2, 3)NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.IOPEP-02.2.1 I Rev. 6 Page 46 of 310 ATTACHMENT 1 Page 21 of 219 EAL Bases 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 ICRG1.BNP Basis Reference(s):

1. EMG-NGGC-0002 Off-site Dose Assessment

2. OPEP-03.4.7 Automation of Off-Site Dose Projections

3. OE&RC-03.4.8, Offsite Dose Projections for Monitored Releases 4. NEI 99-01 AS1{Deleted:

AG1 IOPEP-02.2.1 R Rev. 6 Page 47 of 310 ATTACHMENT 1 Page 22 of 219 EAL Bases Category: Subcategory: Initiating Condition: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL: RS1.3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: " Closed window dose rates> 100 mR/hr expected to continue for > 60 min.* Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation.(Notes 1,2)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Mode Applicability: All Definition(s): SITE BOUNDARY-Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan BNP Basis: OPEP-02.6.6, Environmental Monitoring Team Leader and OPEP-03.5.5 Environmental Monitoring and Plume Tracking provide guidance for emergency or post-accident radiological environmental monitoring (ref. 1, 2).IOPEP-02.2.1 I Rev. 6 1 Page 48 of 310 ATTACHMENT 1 Page 23 of 219 EAL Bases NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that 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.Escalation of-the emergency classification level would be via IC.RG1.BNP Basis Reference(s):

1. OPEP-02.6.6, Environmental Monitoring Team Leader 2. 0PEP-03.5.5 Environmental Monitoring and Plume Tracking-Deleted: Classification based on effluent-.-.--.-.--

.- .--, -monitor readings assumes that a release pathl to the environment is established. If the effluent flow past an effluent monitor is known to have due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.¶ { Deleted: AG1 3. NEI 99-01 AS1 IOPEP-02.2.1 I Rev. 6 Page 49 of 310 ATTACHMENT 1 Page 24 of 219 EAL Bases R -Abnormal Rad Levels / Rad Effluent Category: Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL: RG1.1 General Emergency In the absence of real-time dose assessment, reading on any Table R-1 effluent radiation monitor > column "GE" for > 15 min.(Notes 1,2, 3, 4)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE [ SAE Alert UE Main Stack Rad D12-RM-23S 2.13E+09 pCi/sec 2.13E+08 pCi/sec 2.13E+07 pCi/sec 1.80E+06 pCi/sec Reactor Bldg Vent Noble CAC-AQH-1264-3Gas E Turbine Bldg Vent D12-RM-23 1.07E+08 pCi/sec 1.07E+07 pCi/sec 1.07E+06 pCi/sec 1.13E+04 p.Ci/sec Service Water Effluent D12-RM-K605

.....

2 x hi alarm Rad Radwaste Effluent Rad D12-RM-K604

---- .... 2 x hi-hi alarm I OPEP-02.2.1 Rev. 6 1 Page 50 of 310 ATTACHMENT 1 Page 25 of 219 EAL Bases Mode Applicability:

All Definition(s): None BNP Basis: This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either (ref. 2): 0 1000mRemTEDE

5000 mRem CDE Thyroid The column "SAE" gaseous effluent release values in Table R-1 correspond to calculated doses of 100% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid).Instrumentation that may be used to assess this EAL is listed below (ref 1):* Main Stack Monitoring System Noble Gas Activity Monitor 2-D12-RM-23S (1/2-D12-RR-4599-4)* Turbine Building Ventilation Monitoring System Noble Gas Activity Monitor 1(2)-D12-RM-23 (1/2-DL12-RR-4548-4)

NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.I OPEP-02.2.1 I Rev. 6 1 Page 51 of 310 ATTACHMENT 1 Page 26 of 219 EAL Bases 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.BNP Basis Reference(s):

1. BNP ODCM Appendix E Radioactive Liquid and Gaseous Effluent Monitoring Instrumentation Numbers 2. EP-EALCALC-BNP-0801 Radiological Gaseous Effluent Values (EALs RG1, RS1, RA1 and RU1)3. NEI 99-01 AG1 IOPEP-02.2.1 IRev. 6 1Page 52 of 310 ATTACHMENT 1 Page 27 of 219 EAL Bases R -Abnormal Rad Levels / Rad Effluent Category: Subcategory:

1 -Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL: RG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 3, 4)Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RS1I.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All Definition(s): SITE BOUNDARY-Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan BNP Basis: Dose assessments are performed by computer-based methods (ref. 1, 2, 3)NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.I OPEP-02.2.1 Rev. 6 1 Page 53 of 310 ATTACHMENT 1 Page 28 of 219 EAL Bases 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.BNP Basis Reference(s):

1. EMG-NGGC-0002 Off-site Dose Assessment

2. 0PEP-03.4.7 Automation of Off-Site Dose Projections

3. OE&RC-03.4.8, Offsite Dose Projections for Monitored Releases 4. NEI 99-01 AG1 IOPEP-02.2.1 IRev. 6 1Page 54 of 310 ATTACHMENT 1 Page 29 of 219 EAL Bases Category: Subcategory:

Initiating Condition: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL: RG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:* Closed window dose rates > 1,000 mR/hr expected to continue for > 60 min.* Analyses of field survey samples indicate thyroid CDE > 5,000 mrem for 60 min. of inhalation.(Notes 1,2)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Mode Applicability: All Definition(s): SITE BOUNDARY-Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan BNP Basis: OPEP-02.6.6, Environmental Monitoring Team Leader and OPEP-03.5.5 Environmental Monitoring and Plume Tracking provide guidance for emergency or post-accident radiological environmental monitoring (ref. 1, 2).IOPEP-02.2.1 I Rev. 6 1 Page 55 of 310 ATTACHMENT 1 Page 30 of 219 EAL Bases NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.B References).

1. OPEP-02.6.6, Environmental Monitoring Team Leader 2. OPEP-03.5.5 Environmental Monitoring and Plume Tracking 3. NEI 99-01 AG1--Deleted: 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.¶ IOPEP-02.2.1 I Rev. 6 Page 56 of 310 ATTACHMENT 1 Page 31 of 219 EAL Bases Category: R -Abnormal Rad Levels / Rad Effluent Subcategory: 2 -Irradiated Fuel Event Initiating Condition: Unplanned loss of water level above irradiated fuel EAL: RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm (A-04 6-6) or indication AND UNPLANNED rise in area radiation levels as indicated by any of the following radiation monitors:* ARM Channel 26 New Fuel Vault* ARM Channel 27 North of Fuel Pool* ARM Channel 28 Between Reactor and Fuel Pool* ARM Channel 29 Cask Wash Area Mode Applicability: All Definition(s): 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.REFUELING PATHWAY-. The reactor refueling cavity, spent fuel pool and fuel transfer canal (cattle chute) comprise the refueling pathway.I OPEP-02.2.1 Rev. 6 1 Page 57 of 310 1 ATTACHMENT 1 Page 32 of 219 EAL Bases BNP Basis: The spent fuel pool low water level alarm setpoint is actuated by level switch G41 O-LSHL-N001 at a setpoint of 37' 5". Water level restoration instructions are performed in accordance with 1(2)APP A-04 6-6 Fuel Pool Level Low (ref. 1).When the spent fuel pool and reactor cavity are connected, there could exist the possibility of uncovering irradiated fuel. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred to and from the RPV and spent fuel pool.NEI 99-01 Basis: 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.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 -------- -Deleted: AA2 BNP Basis Reference(s):

1. 1(2)APP-A-04 6-6 Fuel Pool Level Low 2. DBD-1 1 Radiation Monitoring System 3. NEI 99-01 AU2 OPEP-02.2.1 Rev. 6 Page 58 of 310 ATTACHMENT 1 Page 33 of 219 EAL Bases R -Abnormal Rad Levels / Rad Effluent Category: Subcategory:

2 -Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: RA2.1 Unusual Event Uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability: All Definition(s): REFUELING PATHWAY-. The reactor refueling cavity, spent fuel pool and fuel transfer canal (cattle chute) comprise the refueling pathway.BNP Basis: None.NEI 99-01 Basis: This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool, 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.." Deleted: (see Developer Notes)Deleted: This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HU1. ¶Escalation of the emergency would be based on either Recognition Category A or C ICs.¶EAL #1¶1 IOPEP-02.2.1 I Rev. 6 Page 59 of 310o ATTACHMENT 1 Page 34 of 219 EAL Bases This EAL escalates from RU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations. While an area radiation monitor could detect an 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 ........Escalation of the emergency classification level would be via IC ........... -I Deleted: AU2 I BNP Basis Reference(s):

1. 1(2)APP-A4 6-6 (Fuel pool Level Low)2. 1(2)APP-A7 2-2 (Reactor Water Level Hi/Low)3. NEI 99-01 AA2 I" Deleted: EAL #2T This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly.

A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).¶ EAL #3T 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.W tlL' Deleted: sDeleted: AS1 Deleted: a Deleted: ASi IDeleted: or AS2 (see AS2 Developer Notes)IOPEP-02.2.1 I Rev. 6 Page 6oof 310o ATTACHMENT 1 Page 35 of 219 EAL Bases R -Abnormal Rad Levels / Rad Effluent Category: Subcategory: 2 -Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND Any of the following radiation monitor indications: " Reactor Bldg Vent Rad Monitor Channel A or B (> 3 mR/hr)" ARM Channel 26 New Fuel Vault (> 6 mR/hr)" ARM Channel 27 North of Fuel Pool (>10 mR/hr)" ARM Channel 28 Between Reactor and Fuel Pool (> 1000 mR/hr)* ARM Channel 29 Cask Wash Area (>40 mR/hr)Mode Applicability: All Definition(s): None BNP Basis: The high alarm setpoints for the radiation monitors are (ref. 1, 2, 3 4): " Reactor Building Exhaust Plenum Rad Monitor Channel A or B > 3 mR/hr" ARM Channel 26 New Fuel Vault > 6 mR/hr" ARM Channel 27 North of Fuel Pool > 10 mR/hr" ARM Channel 28 Between Reactor and Fuel Pool > 1000 mR/hr* ARM Channel 29 Cask Wash Area > 40mR/hr I OPEP-02.2.1 I Rev. 6 1 Page 61 of 310 ATTACHMENT 1 Page 36 of 219 EAL Bases NEI 99-01 Basis: This IC addresses events that have caused imminent or actual damage to an irradiated fuel I 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.--- Deleted: (see Developer Notes)gscalation of-the emergency would be-based on-either Recognition Category_ or CICs.This EAL addresses a release of radioactive material caused by mechanical damaqe 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).. Escalation of the emergency classification level would be via IQDS1 .BNP Basis Reference(s):

1. 1(2)APP-UA-03 3-7 2. 1(2)APP-UA-03 4-5 3. 1(2)APP-UA-03 4-7 1 II II I'II II II 4.5.DBD-1 1 Radiation Monitoring System NEI 99-01 AA2 Deleted: This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HU1. ¶Deleted: A Deleted: EAL #This EAL escalates from AU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters.

Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations. ¶While an area radiation monitor could detect an 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.¶Deleted: EAL #3Spent 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.Deleted: As Deleted: ASI Deleted: or AS2 (see AS2 Developer Notes)I OPEP-02.2.1 Rev. 6 1 Page 62 of 310 ATTACHMENT 1 Page 37 of 219 EAL Bases R -Abnormal Rad Levels / Rad Effluent Category: Subcategory: 2 -Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: RA2.3 Alert Lowering of spent fuel pool level to < 105 ft. 3 in. ele.Mode Applicability: All Definition(s): None BNP Basis: Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1 -116 ft. 1 in. ele.), SFP level 10 ft. above the top of the fuel racks (Level 2 -105 ft. 3 in. ele.) and SFP level at the top of the fuel racks (Level 3-95 ft. 3 in. ele.).An indicated level of 105 ft. 3 in. corresponds to the Level 2 setpoint (Ref. 1).OPEP-02.2.1 Rev. 6 Page 63 of 310 ATTACHMENT 1 Page 38 of 219 EAL Bases NEI 99-01 Basis: This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool, _These-events

-Deleted: (see Developer Notes)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.Escalation of the emergency would be-based on-either Recognition Cate9gory_,orC Cs. _---------------------------------------------------,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 IQRS1 BNP Basis Reference(s):

1. PCHG-DESG Engineering Change 0000089578R0

2. NEI 99-01 AA2 II II'I II 51 II Si 5'SI Ii II II ii Ii SI SI SI Iii Iii nil nil III Deleted: This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HUI. T Deleted: A Deleted: EAL #This EAL escalates from AU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters.

Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations. T 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.¶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). Deleted: EAL #3 Deleted: a Deleted: AS1 Deleted: or AS2 (see AS2 Developer Notes)J IOPEP-02.2.1 I Rev. 6 Page 64 of 310 ATTACHMENT 1 Page 39 of 219 EAL Bases Category: R -Abnormal Rad Levels / Rad Effluent Subcategory: 2 -Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL: RS2.1 Site Area Emergency Lowering of spent fuel pool level to < 95 ft. 3 in. ele.Mode Applicability: All Definition(s): None BNP Basis: Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1 -116 ft. 1 in. ele.), SFP level 10 ft. above the top of the fuel racks (Level 2 -105 ft. 3 in. ele.) and SFP level at the top of the fuel racks (Level 3-95 ft. 3 in. ele.).An indicated level of 95 ft. 3 in. corresponds to the Level 3 setpoint (Ref. 1).NEI 99-01 Basis: This EAL addresses a significant loss of spent fuel pool inventory control and makeup --- -- Deleted: IC 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 AG1 or G2. Deleted: AG2 BNP Basis Reference(s):

1. PCHG-DESG Engineering Change 0000089578R0 OPEP-02.2.1 Rev. 6 Page 65 of 310

2. NEI 99-01 AS2 OPEP-02.2.1 ev. 61 Page 66 of 310 ATTACHMENT 1 Page 41 of 219 EAL Bases Category:

R -Abnormal Rad Levels / Rad Effluent Subcategory: 2 -Irradiated Fuel Event Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL: RG2.1 General Emergency Spent fuel pool level cannot be restored > 95 ft. 3 in. ele. for > 60 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: All Definition(s): None BNP Basis: Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1 -116 ft. 1 in. ele.), SFP level 10 ft. above the top of the fuel racks (Level 2 -105 ft. 3 in. ele.) and SFP level at the top of the fuel racks (Level 3-95 ft. 3 in. ele.).An indicated level of 95 ft. 3 in. corresponds to the Level 3 setpoint (Ref. 1).NEI 99-01 Basis: This EAL addresses a significant loss of sprent fuel pool inventory control and makeup-------- -Deleted: IC 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. BNP Basis Reference(s):

1. PCHG-DESG Engineering Change 0000089578R0 OPEP-02.2.1 Rev. 6 Page 67 of 310

2. NEI 99-01 AG2 I OPEP-02.2.1 I Rev. 6 1 Page 68 of 310 ATTACHMENT 1 Page 43 of 219 EAL Bases R -Abnormal Rad Levels / Rad Effluent Category: Subcategory:

3 -Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL: RA3.1 Alert Dose rates > 15 mR/hr in EITHER of the following areas: Control Room (ARM Channel 1-1)OR Central Alarm Station (by survey)Mode Applicability: All Definition(s): None.- --Formatted: Line spacing: single BNP Basis: Areas that meet this threshold include the Control Room and the Central Alarm Station (CAS).ARM Channel 1-1 monitors the Control room for area radiation (ref. 1). The CAS is included in this EAL because of its' importance to permitting access to areas required to assure safe plant operations. There is no permanently installed CAS area radiation monitors that may be used to assess this EAL threshold. Therefore this threshold must be assessed via local radiation survey for the CAS.OPEP-02.2.1 Rev. 6 Page 69 of 310 ATTACHMENT 1 Page 44 of 219 EAL Bases NEI 99-01 Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. TheSite Emergency

-- Deleted: Emergency Director Coordinator should consider the cause of the increased radiation levels and determine if anotheri C maye e applicablea

1 I ,Escalation of -the emergency classification level would be via Recognition CatqgorYB, C or F ICs.BNP Basis Reference(s):

1. 1 (2)APP-UA-03 6-7 (Area RAID Control Room Hi)2. NEI 99-01 AA3 Deleted: For EAL #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the 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).Deleted: 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 roonVarea (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. T<#>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.¶Deleted: A IOPEP-02.2.1 I Rev. 6 Page 70 of 310 ATTACHMENT 1 Page 45 of 219 EAL Bases R -Abnormal Rad Levels / Rad Effluent Category: Subcategory: 3 -Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL: RA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table R-2 rooms or areas (Note 5)Note 5: 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. Table R-2 Safe Operation & Shutdown Areas Room/Area Mode Applicability Reactor Building -17' North RHR Unit-1 & 2 3, 4, 5 Reactor Building -17' South RHR Unit-1 & 2 3, 4, 5 Reactor Building 20' East & West MCC Areas Unit-1 & 2 3, 4, 5 Reactor Building 20' Pipe Tunnel Unit-1 & 2 3, 4, 5 Mode Applicability: All Definition(s): IMPEDE(D)- Personnel access to a room or area is hindered to an extent that 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). 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.BNP Basis: If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.OPEP-02.2.1 Rev. 6 Page 71 of 310 ATTACHMENT 1 Page 46 of 219 EAL Bases The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).NEI 99-01 Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. ThegSite Emergency -.--- --- Deleted: Emergency Director Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable. For RA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, -Deleted: EAL#2 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: " 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.IOPEP-02.2.1 I Rev. 6 1 Page 72 of 310 ATTACHMENT 1 Page 47 of 219 EAL Bases If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.Escalation of the emergency classification level would be via Recognition Category., C or F -ICs.BNP Basis Reference(s):

1. Attachment 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases 2. NEI 99-01 AA3 , Deleted: A I OPEP-02.2.1 I Rev. 6 1 Page 73 of 310 ATTACHMENT 1 Page 48 of 219 EAL Bases Category C -Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature < 200 0 F); EALs in this category are applicable only in one or more cold operating modes.Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, containment closure, and fuel clad integrity for the applicable operating modes (4 -Cold Shutdown, 5 -Refueling, D -Defueled). The events of this category pertain to the following subcategories:

1. RPV Level Reactor Pressure Vessel water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.

2. Loss of Emergency AC Power Loss of essential plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

This category includes loss of onsite and offsite power sources for 4160 VAC emergency buses.3. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.

4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

This category includes loss of power to or degraded voltage on the 125 VDC vital buses.ATTACHMENT 1 OPEP-02.2.1 Rev. 6 Page 74 of 310 Page 49 of 219 EAL Bases 5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

6. Hazardous Event Affectino Safety Systems Certain hazardous natural and technological events may result in visible damage to or degraded performance of safety systems warranting classification.

IOPEP-02.2.1 I Rev. 1 Page 75 of 310 ATTACHMENT 1 Page 50 of 219 EAL Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: UNPLANNED loss of RPV inventory for 15 minutes or longer EAL: CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RPV water level less than a required lower limit for _ 15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): 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.BNP Basis: Figure C-1 illustrates the elevations of the RPV level instrument ranges (ref. 2).With the plant in Cold Shutdown, RPV water level is normally maintained above the RPV low level scram setpoint of 166 in. above TAF (ref. 1, 3). However, if RPV level is being controlled below the RPV low level scram setpoint, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern.With the plant in Refueling mode, RPV water level is normally maintained at or above the reactor vessel flange (Technical Specification LCO 3.9.6 requires at least 23 ft of water above the top of the reactor vessel flange in the refueling cavity during refueling operations). The RPV flange is at an indicated level of 355 in. as indicated on the red scale of B21-LI-R605AIB Shutdown Range Reactor Water Level Indication (ref. 4).IOPEP-02.2.1 I Rev. 6 Page 76 of 3101 ATTACHMENT 1 Page 51 of 219 EAL Bases NEI 99-01 Basis: This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor PPV 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.--Deleted: (reactor vesse[/RCS [PWR] or Deleted: [BWRF)This EAl, recognizes that the minimum required ,RPVlevel can chanqe several times durin~g_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.Continued loss of RCS inventory may result in escalation to the Alert emergency classification --level via either IC CAl or CA3.BNP Basis Reference(s):

1. OEOP-01-NL EOP-SAMG NUMERICAL LIMITS AND VALUES, Table 1E 2. SD-01.2 Reactor Vessel Instrumentation Figure 01.2-1 Reactor Water Level Instrument Ranges 3. 1(2) APP A7 2-2 (Reactor Water Level Hi/Low)4. OGP-06 Cold Shutdown to Refueling (Head Unbolted) step 5.1.14 5. NEI 99-01 CUl Deleted: #1 Deleted: (reactor vessel/RCS

[PWR] or 1 Deleted: [BWR])Deleted: EAL #2 addresses a condition where all means to determine (reactor vessel/RCS (PWR] or RPV [BWR]) 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 (reactor vesseV/RCS [PWR] or RPV [BWR]).¶IOPEP-02.2.1 I Rev. 6 Page 77 of 310 ATTACHMENT 1 Page 52 of 219 EAL Bases Figure C-1 RPV Levels (ref. 2)Reactor Water Level Instrument Ranges 917" NPPROX 850' FROM VESSEL ZERO 4PRO 850 I OPEP-02.2.1 I Rev. 6 1 Page 78 of 310 ATTACHMENT 1 Page 53 of 219 EAL Bases C -Cold Shutdown / Refueling System Malfunction Category: Subcategory: 1 -RPV Level Initiating Condition: UNPLANNED loss of RPV inventory EAL: CU1.2 Unusual Event RPV water level cannot be monitored AND UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory Table C-1 Sumps & Tanks* Drywell Floor Drain Sump* Drywell Equipment Drain Sump* RB Floor Drain Sump* RB Equipment Drain Sump* Torus* Visual Observation Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): 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.I OPEP-02.2.1 Rev. 6 1 Page 79 of 310 ATTACHMENT 1 Page 54 of 219 EAL Bases BNP Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refuel mode is normally monitored using the red scale of B21 -LI-R605A/B Shutdown Range Reactor Water Level Indication. In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (ref. 1). A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage (ref. 5). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. NEI 99-01 Basis: This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitorlPV level concurrent with -_ Deleted: (reactor vessel/RCS [PWR] or indications of coolant leakage. Either of these conditions is considered to be a potential Deleted: [BWR])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..ThsEAL,_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 Deleted: EAL #1 recognizes that the minimum required (reactor vesseVRCS [PWR] or RPV[BWR]) 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 ff 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.¶I Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.BNP Basis Reference(s):

1. OOP-47 Floor and Equipment Drain System Operating Procedure 2. 101-03.1 Control Room Operator Daily Surveillance Report 0 3. 201-03.2 Control Room Operator Daily Surveillance Report 4. OAOP-14.0 Abnormal Primary Containment Conditions Deleted: #2 D eleted: (reactor vesseVRCS

[PWR] or Deleted: [BWRI)Deleted: (reactor vessel/Deleted: [PWR] or RPV [BWR])IOPEP-02.2.1 I Rev. 6 1 Page 80 of 310

5. 1(2)OP-17 Residual Heat Removal System Operating Procedure 6. NEI 99-01 CU1 I OPEP-02.2.1 Rev. 6 1 Page 81 of 310 1 ATTACHMENT 1 Page 56 of 219 EAL Bases C -Cold Shutdown / Refueling System Malfunction Category: Subcategory:

1 -RPV Level Initiating Condition: Loss of RPV inventory EAL: CA1.1 Alert Loss of RPV inventory as indicated by RPV water level < 105 in. above TAF (Level 2)Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): None BNP Basis: The threshold RPV level of 105 in. is the low-low ECCS actuation setpoint (ref. 1). RPV level is normally monitored using the instruments in Figure C-1 (ref. 2).When reactor vessel water level drops to 105 in. above TAF high pressure steam-driven injection sources HPCI (ECCS) and RCIC receive an initiation signal (ref. 1). Although these systems cannot restore RCS inventory in the cold condition, the Low-Low (Level 2) ECCS actuation setpoint is operationally significant and is indicative of a loss of RCS inventory significantly below the low level scram setpoint specified in CU .1.NEI 99-01 Basis: This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.For this EAL, a lowering of water level below. 105 in above TAF, indicates that operator actions Deleted: #1 e ---- --ýf i -------------Deleted: st-p cfclv l have not been successful in restoring and maintainingRPVXwaterlevel. The-heat-uprateof -Delete: (site-specific level)............ ..... ............. e-of Deleted: t the coolant will increase as the available water inventory is reduced. A continuing decrease in -, eltd ft water level will lead to core uncovery. "eee: (reactor vesse/RCS [PWR] or"lDeleted: (1BWR])ATTACHMENT 1 OPEP-02.2.1 Rev. 6 Page 82 of 310 Page 57 of 219 EAL Bases Although related, this EAL,is concerned with the loss of RCS inventory and not the potential .--ue: concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat -. Deleted: Residual Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.j Jf 1 RPV water evel continues to lower, then escalation toSiteAreaEmergency would be via IC --CS1.BNP Basis Reference(s):

1. OEOP-01-NL EOP-SAMG NUMERICAL LIMITS AND VALUES, Table 1E 2. SD-01.2 Reactor Vessel Instrumentation Figure 01.2-1 Reactor Water Level Instrument Ranges* Deleted: For EAL #2, the inability to monitor (reactor vessel/RCS

[PWR] or RPV [BWR])level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels.Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR( or RPV [BWR]).¶i The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CSijI Deleted: the (reactor vessel/RCS [PWRI or Deleted: IBWR]) inventory OPEP-02.2.1 Rev. 6 Page 83 of 310 ATTACHMENT 1 Page 58 of 219 EAL Bases Figure C-1 RPV Levels (ref. 2)Reactor Water Level Instrument Ranges 917" p FROM VESSEL ZERO APPROX 850'H T 0-NORMAL LEVEL-TAF (-8.44")LI:E M.gI W STEAM UNE N 4- SPILL-OVER-b IS 25 ir ON R N027 A N NARROW a RANGE E W I 0 E N004 N027 R A N 0 E N02 F U E L z ---D 0 N F N03SIN037 I OPEP-02.2.1 Rev. 6 Page 84 of 310 ATTACHMENT 1 Page 59 of 219 EAL Bases C -Cold Shutdown / Refueling System Malfunction Category: Subcategory: 1 -RPV Level Initiating Condition: Loss of RPV inventory EAL: CA1.2 Alert RPV water level cannot be monitored for -15 min. (Note 1)AND UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory Table C-1 Sumps & Tanks* Drywell Floor Drain Sump* Drywell Equipment Drain Sump* RB Floor Drain Sump* RB Equipment Drain Sump* Torus* Visual Observation Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): 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.BNP Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. In the Refuel mode, the RCS is not intact and RPV level may be monitored by different means, including the ability to monitor level visually.OPEP-02.2.1 Rev. 6 Page 85 of 310 ATTACHMENT 1 Page 60 of 219 EAL Bases In this EAL, all water level indication would be unavailable, and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Sump level increases must be evaluated against other potential sources of leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (ref. 1 thru 4). A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV.With RHR System operating in the Shutdown Cooling mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage (ref. 5). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. NEI 99-01 Basis: This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.,=or this EAL,_the inability to monitorRPVjlevel maybe caused byinstrumentation 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 'fro m th e P CS ..-- - .................................................The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1.If the Rq4 inventory level continues to lower, then escalation to Site Area Emergency would ....Deleted: For EAL #1, a lowering of water level below (site-specific level) indicates that operator actions have not been successful in restoring and maintaining (reactor vesseVRCS[PWR] or RPV [BWR]) 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 #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.DA via ILI L,u.% % I eet ea I D1 eleted: #2%' I Deleted: (reactor vessel/RCS (PWR] or\\\\Deleted: [BWR])SDeleted: (reactor vessel/Deleted: [PWR] or RPV [BWRI)Deleted: (reactor vessel/Deleted: [PWR] or RPV [BWR])IOPEP-02.2.1 I Rev. 6 Page 86 of 310 ATTACHMENT 1 Page 61 of 219 EAL Bases BNP Basis Reference(s): 1.2.3.4.5.6.OOP-47 Floor and Equipment Drain System Operating Procedure 101-03.1 Control Room Operator Daily Surveillance Report 201-03.2 Control Room Operator Daily Surveillance Report OAOP-1 4.0 Abnormal Primary Containment Conditions 1(2)OP-17 Residual Heat Removal System Operating Procedure NEI 99-01 CA1 I OPEP-02.2.1 I Rev. 6 1 Page 87 of 3101 ATTACHMENT 1 Page 62 of 219 EAL Bases C -Cold Shutdown / Refueling System Malfunction Category: Subcategory: 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1.1 Site Area Emergency CONTAINMENT CLOSURE not established AND RPV level < 45 in. (Level 3)Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): CONTAINMENT CLOSURE -The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to BNP, Containment Closure is established when either Primary Containment is Operable per Section 3.6.1.1 of Technical Specifications or Secondary Containment is considered functional per the requirements of OAP-022, BNP Outage Risk Management. BNP Basis: RPV level is normally monitored using the instruments in Figure C-1 (ref. 2).When RPV level decreases to 45 in., RPV water level is below the low-low-low ECCS actuation setpoint (Level 3) (ref. 1).The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV water level decrease and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier.NEI 99-01 Basis: This IC addresses a significant and prolonged loss ofqCSqinventory control and makeup.... capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS OPEP-02.2.1 Rev. 6 Page 88 of 310-- Deleted: (reactor vessev-4Deleted: [PWR] or FtPV [BWA]) 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 RPV level cannot be restored, fuel damage is probable.--J Deleted: RCS/reactor vessel Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified.RPV levels ofCS1.1 andCS1 .2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. This EAL addresses concerns raised byGeneric 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 CG1 or RGl.....BNP Basis Reference(s):

1. OEOP-01 -NL EOP-SAMG NUMERICAL LIMITS AND VALUES, Table 1 E 2. BNP Technical Specifications, Sections 3.6.1.1--Deleted: R2S/reactor vessel I Deleted: 2.b'I'I Deleted: In EAL 3.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).

It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.¶ The inability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]).¶These'JDeleted: a 3.OAP-022, BNP Outage Risk Management, Section 6.5 4. SD-01.2 Reactor Vessel Instrumentation Figure 01.2-1 Reactor Water Level Instrument Ranges 5. NEI 99-01 CS1 Deleted: AG1 IoPEP-02.2.1 I Rev. 6 Page 89 of 310 ATTACHMENT 1 Page 64 of 219 EAL Bases Figure C-1 RPV Levels (ref. 4)Reactor Water Level Instrument Ranges 917" FROM VESSEL ZERO850'S H T JD °0 P ------- 4(*550w-NORMAL LEVEL-TAF CORE L-BAF STEAM UNE 4- SPILL.OVER------ IS 254" ON N027 NARROW RANGE W E U004 R A N G E I N02 W N R A N G E.027 F U E L z <0 N E N038NN037 I OPEP-02.2.1 I Rev. 6 1 Page 90 of 310 ATTACHMENT 1 Page 65 of 219 EAL Bases C -Cold Shutdown / Refueling System Malfunction Category: Subcategory: 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1.2 Site Area Emergency CONTAINMENT CLOSURE established AND RPV level < TAF Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): CONTAINMENT CLOSURE- The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to BNP, Containment Closure is established when either Primary Containment is Operable per Section 3.6.1.1 of Technical Specifications or Secondary Containment is considered functional per the requirements of OAP-022, BNP Outage Risk Management. BNP Basis: When RPV level drops to the top of active fuel (TAF) (an indicated RPV level of --8 in.), core uncovery starts to occur (ref. 1).OPEP-02.2.1 Rev. 6 Page 91 of 310 ATTACHMENT 1 Page 66 of 219 EAL Bases NEI 99-01 Basis:...... -Deleted: (reactor vesset/RCS [PWR] or This IC addresses a significant and prolonged loss of RPVlevel control and makeup capaility Deleted: [r nvenr 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 RPV level ---- Deleted: RCS/reactor vessel cannot be restored, fuel damage is probable.J}J Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of C1.1 and CS1.2 .- Deleted:EALs1.b reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of Deleted: 2.b a fission product release to the environment. JThisEAL_ addresses concerns raised_by Generic Letter 88-17, Loss of Decay Heat Removal; _ -SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1 449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CG1 or RG1.BNP Basis Reference(s):

1. OEOP-01-NL EOP-SAMG NUMERICAL LIMITS AND VALUES 2. BNP Technical Specifications, Sections 3.6.1.1 and 3.6.4.1 3. OAP-022, BNP Outage Risk Management, Section 6.5 II II Deleted: In EAL 3.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).

It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.¶ The inability to monitor (reactor vessel/RCS [PWR] or RPV [BWRJ) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]).¶These Deleted: s Deleted: AG1 IOPEP-02.2.1 I Rev. 6 Page 92 of 310 ATTACHMENT 1 Page 67 of 219 EAL Bases C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Category: Subcategory: Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1.3 Site Area Emergency RPV water level cannot be monitored for > 30 min. (Note 1)AND Core uncovery is indicated by EITHER of the following: " UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory" UNPLANNED increase in ARM Channel 28 Between Reactor and Fuel Pool> 1000 mR/hr Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table C-1 Sumps & Tanks* Drywell Floor Drain Sump* Drywell Equipment Drain Sump* RB Floor Drain Sump* RB Equipment Drain Sump* Torus* Visual Observation Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): 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.IOPEP-02.2.1 I Rev. 6 Page 93 of 310 ATTACHMENT 1 Page 68 of 219 EAL Bases BNP Basis: If RPV level monitoring capability is unavailable, the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Sump level increases must be evaluated against other potential sources of leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (ref. 1 thru 4). A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage (ref. 5). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory. In the Refueling Mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in indications on installed area radiation monitors. ARM Channel 28 Between Reactor and Fuel Pool is located on the Refuel Floor in the Reactor Building and is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations. If this radiation monitor reaches and exceeds the alarm setpoint of 1,000 mr/hr, a loss of inventory with potential to uncover the core is likely to have occurred (Ref. 6, 7).NEI 99-01 Basis: This IC addresses a significant and prolonged loss of ,eactor vessel/RCS inventory control Deleted: and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a --Deleted: [PWR] or RPV [BWR])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. IfRPV level cannot be restored, fuel damage is probable.--Deleted: RCS/reactor vessel ,he 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 monitorPV ._level-may be caused by instrumentation and/orpower 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 Deleted: Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of EALs 1 .b and 2.b reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment,¶ In EAL 3.a, t Deleted: (reactor vesseVRCS [PWR] or Deleted: [BWR])I OPEP-02.2.1 I Rev. 6 1 Page 94 of 310 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 Deleted: (reactor vessel/" Deleted: [PWR] or RPVaddLresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; _-- 4 Deleted: These BWRI)SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown " el-ed:s 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 CG1 or RG1 , I Deleted: AG1 I I OPEP-02.2.1 I Rev. 6 1 Page 95 of 3101 ATTACHMENT 1 Page 70 of 219 EAL Bases BNP Basis Reference(s):

1. OOP-47 Floor and Equipment Drain System Operating Procedure 2. 101-03.1 Control Room Operator Daily Surveillance Report 3. 201-03.2 Control Room Operator Daily Surveillance Report 4. OAOP-14.0 Abnormal Primary Containment Conditions

5. 1(2)OP-17 Residual Heat Removal System Operating Procedure 6. 1(2)APP-UA-03 4-7 7. DBD-1 1 Radiation Monitoring System 8. NEI 99-01 CS1 IOPEP-02.2.1 ev. 61 Page 96 of 310 ATTACHMENT 1 Page 71 of 219 EAL Bases Category: Subcategory:

Initiating Condition: EAL: C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Loss of RPV inventory affecting fuel clad integrity with Containment challenged CG1.1 General Emergency RPV level < TAF for > 30 min. (Note 1)AND Any Containment Challenge indication, Table C-2 Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.Table C-2 Containment Challenge Indications

CONTAINMENT CLOSURE not established (Note 6)* Primary Containment hydrogen concentration

> 6%* UNPLANNED rise in PC pressure* Exceeding one or more Secondary Containment Control Maximum Safe Operating Area Radiation Levels (OEOP-03-SCCP Table 3)Mode Applicability: 4 -Cold Shutdown, 5 -Refuel I OPEP-02.2.1 I Rev. 6 1 Page 97 of 310 ATTACHMENT 1 Page 72 of 219 EAL Bases Definition(s): CONTAINMENT CLOSURE -The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to BNP, Containment Closure is established when either Primary Containment is Operable per Section 3.6.1.1 of Technical Specifications or Secondary Containment is considered functional per the requirements of OAP-022, BNP Outage Risk Management. BNP Basis: When RPV level drops below --8 in., the top of active fuel, core uncovery starts to occur (ref.6).Four conditions are associated with a challenge to Primary Containment (PC) integrity:

CONTAINMENT CLOSURE is not established (Ref. 7)." 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 mixture of dissolved gases in the Primary Containment.

However, Primary Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. An explosive mixture can be formed when hydrogen gas concentration in the Primary Containment atmosphere is greater than 6%by volume in the presence of oxygen (>5%) (ref. 2, 3). In Cold Shutdown and Refueling modes it is assumed that the Primary Containment is de-inerted. Monitors CAC-AT-4409 and 4410 monitor hydrogen and oxygen concentration in the drywell during normal operation. These monitors are isolated by accident isolation signals. However, monitors CAC-AT-4409 and 4410 will be realigned to the primary containment for post-accident monitoring via an operator actuated isolation signal override circuit when directed by the EOPs. (ref. 4)." Any unplanned increase in PC pressure in the Cold Shutdown or Refueling mode indicates a potential loss of containment closure capability. Unplanned Primary Containment pressure increases indicates containment closure cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release." Secondary Containment radiation monitors should provide indication of increased release that may be indicative of a challenge to Containment Closure. The Maximum Safe Operating Values are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in OEOP-03-SCCP, Secondary Containment Control, (ref. 5).I0PEP-02.2.1 I Rev. 6 1 Page 98 of 310 1 ATTACHMENT 1 Page 73 of 219 EAL Bases NEI 99-01 Basis: 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 BPV level cannot be restored, fuel damage is probable.With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. The 30-minute criterion is tied to a readily recognizable event start time (.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. _addresses concerns raised_byGeneric 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. 0PEP-02.2.1 IRev. 6 1Page 99 of 310--J Deleted: RCS/reactor vessel I , J Deleted: In EAL 2.b, t xl Deleted: The inability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWRq or RPV [BWR]).¶Deleteds Deleted: a ATTACHMENT 1 Page 74 of 219 EAL Bases BNP Basis Reference(s): 1.2.3.4.5.6.7.8.BNP Technical Specifications Sections 3.6.1.1 and 3.6.4.1 BWROG EPG/SAG Revision 2, Sections PC/G OEOP-02-PCCP, Primary Containment Control Updated FSAR section 6.2.5.2.2 OEOP-03-SCCP, Secondary Containment OEOP-01-NL EOP-SAMG NUMERICAL LIMITS AND VALUES OAP-022 BNP Outage Risk Management, Section 6.5 NEI 99-01 CG1 OPEP-02.2.1 Rev. 6 Page 100 of 310 ATTACHMENT 1 Page 75 of 219 EAL Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with Containment challenged EAL: CG1.2 General Emergency RPV level cannot be monitored for > 30 min. (Note 1)AND Core uncovery is indicated by EITHER of the following: " UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory" UNPLANNED increase in ARM Channel 28 Between Reactor and Fuel Pool > 1000 mR/hr AND Any Containment Challenge indication, Table C-2 Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.Table C-1 Sumps & Tanks* Drywell Floor Drain Sump* Drywell Equipment Drain Sump* RB Floor Drain Sump* RB Equipment Drain Sump* Torus* Visual Observation I OPEP-02.2.1 I Rev. 6 1 Page 101 of 310 ATTACHMENT 1 Page 76 of 219 EAL Bases Table C-2 Containment Challenge Indications" CONTAINMENT CLOSURE not established (Note 6)* Primary Containment hydrogen concentration > 6%* UNPLANNED rise in PC pressure" Exceeding one or more Secondary Containment Control Maximum Safe Operating Area Radiation Levels (OEOP-03-SCCP Table 3)Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): CONTAINMENT CLOSURE -The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to BNP, Containment Closure is established when either Primary Containment is Operable per Section 3.6.1.1 of Technical Specifications or Secondary Containment is considered functional per the requirements of OAP-022, BNP Outage Risk Management. 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.BNP Basis: If RPV level monitoring capability is unavailable, the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Sump level increases must be evaluated against other potential sources of leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (ref. 1 thru 4). A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage (ref. 5). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. OPEP-02.2.1 Rev. 6 Page 102 of 310 In the Refueling Mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in indications on installed area radiation monitors. ARM Channel 28 Between Reactor and Fuel Pool is located on the Refuel Floor in the Reactor Building and is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations. If this radiation monitor reaches and exceeds the alarm setpoint of 1,000 mr/hr, a loss of inventory with potential to uncover the core is likely to have occurred (Ref. 6, 7).Four conditions are associated with a challenge to Primary Containment (PC) integrity: " CONTAINMENT CLOSURE is not established (Ref. 13)." 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 mixture of dissolved gases in the Primary Containment. However, Primary Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. An explosive mixture can be formed when hydrogen gas concentration in the Primary Containment atmosphere is greater than 6%by volume in the presence of oxygen (>5%) (ref. 9, 10). In Cold Shutdown and Refueling modes it is assumed that the Primary Containment is de-inerted. Monitors CAC-AT-4409 and 4410 monitor hydrogen and oxygen concentration in the drywell during normal operation. These monitors are isolated by accident isolation signals. However, monitors CAC-AT-4409 and 4410 will be realigned to the primary containment for post-accident monitoring via an operator actuated isolation signal override circuit when directed by the EOPs. (ref. 11).* Any unplanned increase in PC pressure in the Cold Shutdown or Refueling mode indicates a potential loss of containment closure capability. Unplanned Primary Containment pressure increases indicates containment closure cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release." Secondary Containment radiation monitors should provide indication of increased release that may be indicative of a challenge to Containment Closure. The Maximum Safe Operating Values are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in OEOP-03-SCCP, Secondary Containment Control, (ref. 12).OPEP-02.2.1 Rev. 6 Page 103 of 310 ATTACHMENT 1 Page 78 of 219 EAL Bases NEI 99-01 Basis: 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 RPV level ----- Deleted: RCS/reactor vessel cannot be restored, fuel damage is probable.With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. _The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of --Deleted: In EAL 2.b, 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 _PVVlevelmay be caused by instrumentation and/orpower failures, or .-Deleted: (reactor vessel/RCS [PWR] or water level dropping below the range of available instrumentation. If water level cannot be Deleted: [BWRJ)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 theRCS .-.-.-..... ...Deleted: (reactor vessel/tDeleted: [PWR] or RPV [BWRJ)I0PEP-02.2.1 Rev. 1 Page 104 of 310 ATTACHMENT 1 Page 79 of 219 EAL Bases ThiEAL.addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal;- Deleted: ese SECY91--283,Evaluation-of Shu-tdown andLow-Power Risk/Issues;-INUR-EG-1449, Shutdown Deleted: s 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. BNP Basis Reference(s):

1. OOP-47 Floor and Equipment Drain System Operating Procedure 2. 101-03.1 Control Room Operator Daily Surveillance Report 3. 201-03.2 Control Room Operator Daily Surveillance Report 4. OAOP-14.0 Abnormal Primary Containment Conditions

5. 1(2)OP-17 Residual Heat Removal System Operating Procedure 6. 1(2)APP-UA-03 4-7 7. DBD-1 1 Radiation Monitoring System 8. BNP Technical Specifications Section 3.6.1.1 and 3.6.4.1 9. BWROG EPG/SAG Revision 2, Sections PC/G 10. OEOP-02-PCCP Primary Containment Control 11. Updated FSAR section 6.2.5.2.2 12.OEOP-03-SCCP Secondary Containment Control 13. OAP-022 BNP Outage Risk Management, Section 6.5 14. NEI 99-01 CG1 OPEP-02.2.1 Rev. 6 Page 105 of 310 ATTACHMENT 1 Page 80 of 219 EAL Bases Category: Subcategory:

Initiating Condition: C -Cold Shutdown / Refueling System Malfunction 2 -Loss of Emergency AC Power Loss of all but one AC power source to emergency buses for 15 minutes or longer EAL: CU2.1 Unusual Event AC power capability to Emergency 4 KV Buses El (E3) and E2(E4) reduced to a single power source for > 15 min. (Note 1)AND Any additional single power source failure will result in loss of all unit-specific AC power to SAFETY SYSTEMS Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 4 -Cold Shutdown, 5 -Refuel, D -Defueled Definition(s): 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 (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. I OPEP-02.2.1 Rev. 6 1 Page 106 of 310 ATTACHMENT 1 Page 81 of 219 EAL Bases BNP Basis: The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. 4160 VAC power is divided into Balance of Plant (BOP Bus), and Emergency (E-Bus) distribution. The BOP Buses consist of the Conventional Buses (Common A/B Buses;Buses 1 B/2B, 1 C/2C, 1 D/2D) and Caswell Beach Buses A/B. The Emergency switchgear are Buses El/E3 (Division I) and E2/E4 (Division II) (Ref. 1, 2).The E-Buses are normally powered through the respective BOP Buses (1 D to El, 1 C to E2, 2D to E3, 2C to E4) via a master/slave breaker arrangement. Each E-Bus has a dedicated Diesel Generator to supply an emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of offsite power. The DGs will automatically start and tie onto the E-Buses if the normal power source or offsite power is lost. The DGs can provide power to the E-Buses only. In the event the diesel generator is unavailable for an E-Bus, crosstie capability exists for each E-Bus from the same division of the opposite unit (El to E3, E2 to E4). Although the E-Buses within the unit also have crosstie capability, this alignment is not permitted by plant procedures, with the exception of El to E2 during specific Alternate Safe Shutdown (ASSD) conditions. During periods of unit shutdown, when the Startup Auxiliary Transformer (SAT) would be the only normal source of offsite power, the Unit Auxiliary Transformer (UAT) can be made available by establishing a UAT backfeed. Backfeed from the UAT will require the use of keys for the control selector switches and opening of the respective generator's manual no-load disconnect (Ref. 3).Because 2 RHR pumps on each unit are powered from the unaffected unit, the words "unit-specific" have been added to clarify that the cross-connected RHR pump power cannot be credited as an AC power source relative to this EAL.This cold condition EAL is equivalent to the hot condition EAL SA1.1.NEI 99-01 Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as 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.0PEP-02.2.1 I Rev. 6 1 Page 107 of 310 ATTACHMENT 1 Page 82 of 219 EAL Bases An "AC power source" is a source recognized in AOPs_ and capable of supplying reguired power to an g~g~ency bus. Some examples of this condition-are presented below." A loss of all offsite power with a concurrent failure ofone division of emergencypower _sources (e.g., ,nsite diesel generators)-.-................................." A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator." A loss of emergency power sources (e.g., onsite diesel generators) with a single division of emergency 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.BNP Basis Reference(s):

1. Drawing BN-50.0.01 Electrical Distribution

2. BNP Updated FSAR Chapter 8 3. 1(2)OP-50 Plant Electric System Operating Procedure 4. 1(2)APP-UA15 2-1 (Bus E-1 Undervoltage)

5. 1(2)APP-UA1 6 2-1 (Bus E-2 Undervoltage)

6. 1(2)APP-UA1 7 2-1 (Bus E-3 Undervoltage)

7. 1(2)APP-UA18 2-1 (Bus E-4 Undervoltage)

8. NEI 99-01 CU2-Deleted: and EOPs--Deleted: essential-Deleted: alt but--Deleted: an--Deleted: train I OPEP-02.2.1 Rev. 6 1 Page 108 of 310 ATTACHMENT 1 Page 83 of 219 EAL Bases Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 2 -Loss of Emergency AC Power Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer EAL: CA2.1 Alert Loss of all offsite and all onsite AC power capability to Emergency 4 KV Buses El (E3)and E2(E4) for _ 15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 4 -Cold Shutdown, 5 -Refuel, D -Defueled BNP Basis: The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. 4160 VAC power is divided into Balance of Plant (BOP Bus), and Emergency (E-Bus) distribution. The BOP Buses consist of the Conventional Buses (Common A/B Buses;Buses 1 B/2B, 1 C/2C, 1 D/2D) and Caswell Beach Buses A/B. The Emergency switchgear are Buses El/E3 (Division I) and E2/E4 (Division II) (Ref. 1,2).The E-Buses are normally powered through the respective BOP Buses (1 D to El, 1 C to E2, 2D to E3, 2C to E4) via a master/slave breaker arrangement. Each E-Bus has a dedicated Diesel Generator to supply an emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of offsite power. The DGs will automatically start and tie onto the E-Buses if the normal power source or off site power is lost. The DGs can provide power to the E-Buses only. In the event the diesel generator is unavailable for an E-Bus, crosstie capability exists for each E-Bus from the same division of the opposite unit (El to E3, E2 to E4). Although the E-Buses within the unit also have crosstie capability, this alignment is not permitted by plant procedures, with the exception of El to E2 during specific Alternate Safe Shutdown (ASSD) conditions. During periods of unit shutdown, when the Startup Auxiliary Transformer (SAT) would be the only normal source of offsite power, the Unit Auxiliary Transformer (UAT) can be made available by establishing a UAT backfeed. Backfeed from the UAT will require the use of keys for the control selector switches and opening of the respective generator's manual no-load disconnect (Ref. 3).OPEP-02.2.1 Rev. 6 Page 109 of 310 ATTACHMENT 1 Page 84 of 219 EAL Bases This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EAL SS1.1.NEI 99-01 Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level would be via IC CS1 orRS1 ............ BNP Basis Reference(s):

1. Drawing BN-50.0.01 Electrical Distribution

2. BNP Updated FSAR Chapter 8 3. 1(2)OP-50 Plant Electric System Operating Procedure 4. 1(2)APP-UA15 2-1 (Bus E-1 Undervoltage)

5. 1(2)APP-UA16 2-1 (Bus E-2 Undervoltage)

6. 1(2)APP-UA17 2-1 (Bus E-3 Undervoltage)

7. 1(2)APP-UA18 2-1 (Bus E-4 Undervoltage)

8. NEI 99-01 CA2 , j Deleted: AS1 I IOPEP-02.2.1 IRev. 6 1Page 110 of 310 ATTACHMENT 1 Page 85 of 219 EAL Bases C -Cold Shutdown / Refueling System Malfunction Category: Subcategory:

3 -RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL: CU3.1 Unusual Event UNPLANNED increase in RCS temperature to > 212°F due to loss of decay heat removal capability Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): 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.BNP Basis: Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (212'F, ref. 1). These include (ref.2): " Recirculation Suction Temperatures read on B32-TR-R650 located on panel P-603 (if recirculation loop is in operation)

RHR HX 2A(B) Inlet Temperature as read on E41-TR-R605 Point 1(2), on Panel H12-P614 (RHR HX in service)* RHR HX 2A(B) Outlet Temperature as read on E41-TR-R605 Point 3(4), on Panel H12-P614 (RHR HX not in service)* PPC Display 815, RPV HEATUP/COOLDOWN MONITOR (natural circulation)

NEI 99-01 Basis: This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limitp_..represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the-Site Emergency Coordinator should also refer to IC CA3.OPEP-02.2.1 Rev. 6 1 Page 111 of 310 1 1 Deleted: , or the inability to determine RCS 1 temperature and level, I-j Deleted: Emergency Director ATTACHMENT 1 Page 86 of 219 EAL Bases A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. This EAL involves a loss of decay heat removal capability,or an addition of heat to the RCS in -Delted EAL #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. I During an outage, the level in the reactor vessel will normally be maintained at or 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.I Deleted: EAL #2 Fifteen minutes was selected as a threshold to exclude transient or momentary losses of --there has been aLl indication. instrumentation ca RCS conditions ar to monitor key par Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on core decay heat n exceeding plant configuration-specific time criteria. there is no immed reducedusie th e BNP Basis Reference(s): r since the 1. BNP Technical Specifications Table 1.1-1 2. 1(2)PT-01.7 Heatup/Cooldown Monitoring

3. NEI 99-01 CU3 reflects a condition where significant loss of apability necessary to monitor nd operators would be unable ameters necessary to assure emoval. During this condition, late threat of fuel damage decay heat load has been cessation of power operation.1 IOPEP-02.2.1 Rev.6 Page 112 of 310 ATTACHMENT 1 Page 87 of 219 EAL Bases C -Cold Shutdown / Refueling System Malfunction Category: Subcategory:

3 -RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL: CU3.2 Unusual Event Loss of all RCS temperature and RPV level indication for > 15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): None BNP Basis: RPV water level is normally monitored using the instruments in Figure C-1 (ref. 1).Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (212 0 F, ref. 2). These include (ref.3): " Recirculation Suction Temperatures read on B32-TR-R650 located on panel P-603 (if recirculation loop is in operation)

RHR HX 2A(B) Inlet Temperature as read on E41-TR-R605 Point 1(2), on Panel H12-P614 (RHR HX in service)* RHR HX 2A(B) Outlet Temperature as read on E41-TR-R605 Point 3(4), on Panel H12-P614 (RHR HX not in service)" PPC Display 815, RPV HEATUP/COOLDOWN MONITOR (natural circulation)

NEI 99-01 Basis: I This gEAL addresses jhe inability to determine RCS temperature and RPV level, and Deleted: ic r Deleted: an UNPLANNED increase in RCS represents a potential degradation of the level of safety of the pant. If the RCS is not intact temperature above the Technical Specification and CONTAINMENT CLOSURE is not established during this event, theSite Emergency ...... cold shutdown temperature limit, or Coordinator should also refer to IC CA3. " Deleted: Emergency Director FOPEP-02.2.1 Rev. 6 Page 113 of 310 1 ATTACHMENT 1 Page 88 of 219 EAL Bases This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.BNP Basis Reference(s):

1. SD-01.2 Reactor Vessel Instrumentation Figure 01.2-1 Reactor Water Level Instrument Ranges 2. Technical Specifications Table 1.1-1 3. 1(2)PT-01.7 Heatup/Cooldown Monitoring Deleted: A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.¶ EAL #1 involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications.

During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.¶ During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown. ¶EAL #2 4.NEI 99-01 CU3 I0PEP-02.2.1 I Rev. 6 Page 114 of 310 ATTACHMENT 1 Page 89 of 219 EAL Bases Figure C-1 RPV Levels (ref. 1)Reactor Water Level Instrument Ranges 917" -FROM VESSEL ZERO850'I ( J______________ -.------. f(.FA STEAM UNE*- SPILL.OVER'--) IS 254' ON N027 s H U T O 0 W N R A E E NARROW RANGE-l0E NORMAL LEVEL-TAF O-8.44")LCORE N027 N026 F U E L Z 0 N E NOW35N037 1\1-1 I OPEP-02.2.1 Rev. 6 1 Page 115 of 310 ATTACHMENT 1 Page 90 of 219 EAL Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 3 -RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL: CA3.1 Alert UNPLANNED increase in RCS temperature to > 212'F for > Table C-3 duration (Note 1)OR UNPLANNED RPV pressure increase > 10 psig due to a loss of RCS cooling Note 1: The SEC should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Table C-3: RCS Heat-up Duration Thresholds RCS Status Containment Closure Heat-up Duration Status Intact N/A 60 min.*established 20 min.*Not intact not established 0 min.* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable. Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): CONTAINMENT CLOSURE -The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to BNP, Containment Closure is established when either Primary Containment is Operable per Section 3.6.1.1 of Technical Specifications or Secondary Containment is considered functional per the requirements of OAP-022, BNP Outage Risk Management. OPEP-02.2.1 Rev. 6 Page 116 of 310 ATTACHMENT 1 Page 91 of 219 EAL Bases 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.BNP Basis: A 10 psig RPV pressure increase can be read on (ref. 1): e Indicator PI-R605A located on Panel P603* Indicator PI-R605B located on Panel P601 e Recorder LPR-R608 located on P603* Indicator C32-PI-3332 located on the Remote Shutdown Panel Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (212 0 F, ref. 2). These include (ref.3): " Recirculation Suction Temperatures read on B32-TR-R650 located on panel P-603 (if recirculation loop is in operation)

RHR HX 2A(B) Inlet Temperature as read on E41-TR-R605 Point 1(2), on Panel H12-P614 (RHR HX in service)* RHR HX 2A(B) Outlet Temperature as read on E41-TR-R605 Point 3(4), on Panel H12-P614 (RHR HX not in service)" PPC Display 815, RPV HEATUP/COOLDOWN MONITOR (natural circulation)

NEI 99-01 Basis: This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, The 20-minute criterion --Deleted:, or RCS inventory is reduced (e.g., was included to allow time for operator action to address the temperature increase.e mid-loop operation in PWRs)IOPEP-02.2.1 IRev. 6 1Page 117 of 310 ATTACHMENT 1 Page 92 of 219 EAL Bases The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature 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 --Deleted: or is at reduced inventory [PWfR CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes).This is because 1) the evaporated reactor coolant may be released directly into the Primary Containment or Reactor Building atmosphere and subsequently to the environment, and 2)there is reduced reactor coolant inventory above the top of irradiated fuel.The RFCS ressure increase threshold provides a pressure-based indication of RCS heat-up in Deleted: EAL #2 the. Ahcqtncys rf RCR.. mnnitorinri einahititv_ Escalation of the emergency classification level would be via IC CS1 or RS1.BNP Basis Reference(s):

1. Reactor Vessel Instrumentation System Description SD-011.2 2. BNP Technical Specifications Table 1.1-1 3. 1(2)PT-01.7 Heatup/Cooldown Monitoring

4. Technical Specifications Sections 3.6.1.1 and 3.6.4.1 5. OAP-022, BNP Outage Risk Management

6. NEI 99-01 CA3 Sj- Deleted: AS1 I I OPEP-02.2.1 I Rev. 6 1 Page 118 of 310 ATTACHMENT 1 Page 93 of 219 EAL Bases Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 4 -Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL: CU4.1 Unusual Event< 105 VDC bus voltage indications on Technical Specification required 125 VDC buses for -15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): None BNP Basis: There are two independent divisions per unit, designated Division I and Division II (Batteries 1(2)A-1 and 1(2)A-2 for Division I and Batteries 1(2)B-1 and 1(2)B-2 for Division II) .Each division consists of a 250 VDC battery center tapped to form two 125 VDC batteries. Each 125 VDC battery has an associated full capacity battery charger. The chargers are supplied from the same AC load groups for which the associated DC subsystem supplies the control power.During normal operation, the DC loads are powered from the battery chargers with the batteries floating on the system. In case of loss of normal power to the battery charger, the DC loads are automatically powered from the station batteries. 105 VDC is the minimum design voltage limit (ref. 1).Note that the Control Room DC voltage indicator only reads battery charger output voltage and not battery voltage unless the charger output breaker is closed. However ERFIS does provide DC battery voltage, otherwise battery voltage must be read locally.This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS7.1.I OPEP-02.2.1 I Rev. 6 Page 119 of 310 ATTACHMENT 1 Page 94 of 219 EAL Bases NEI 99-01 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 Division I is out-of-service (inopeprable) for scheduled outage maintenance work and .Division II --Deleted: Train A is in-service (operable), then a loss of Vital DC power affecting Division II would require the Deleted: Train B declaration of an Unusual Event. A loss of Vital DC power to Division I would not warrant an "- 4Deleted: Train B emergency classification. Deleted: Train A Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Depending upon the event, escalation of the emergency classification level would be via IC CAI or CA3, or an IC in Recognition Category,... BNP Basis Reference(s):

1. BNP Technical Specification Bases B.3.8.4 2. OAOP-39.0 LOSS OF DC POWER 3. NEI 99-01 CU4_ -Deleted: A I I OPEP-02.2.1 I Rev. 6 1 Page 120 of 310 ATTACHMENT 1 Page 95 of 219 EAL Bases Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 5 -Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: CU5.1 Unusual Event Loss of all Table C-4 onsite communication methods OR Loss of all Table C-4 Offsite communication methods OR Loss of all Table C-4 NRC communication methods Table C-4 Communication Methods System Onsite Offsite NRC Public Address System X PBX Telephone System X X X Corporate Telephone X X X Communications System Commercial Telephones X X X Satellite Phones X X Cellular Phones X X NRC Emergency X Telecommunications System Mode Applicability: 4 -Cold Shutdown, 5 -Refuel, D -Defueled I OPEP-02.2.1 I Rev. 6 1 Page 121 of 310 ATTACHMENT 1 Page 96 of 219 EAL Bases Definition(s): None BNP Basis: Onsite/offsite/NRC communications include one or more of the systems listed in Table C-4 (ref. 1).Public Address System The Brunswick Plant public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature. This system is powered from the plant uninterruptible power supply which employs battery reserve as well as diesel generator emergency supply.PBX Telephone System The Brunswick Site PBX telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code. The PBX telephone system also provides for outside communications. The PBX switch located in the TSC/EOF building is also backed up by a battery UPS capable of supplying power for a minimum of 8 hours and is augmented by a Diesel Generator capable of supplying power to the TSC/EOF building for at least 5 days.Corporate Telephone Communications System (Voicenet and/or DEMNET)Interconnected through the site PBX and the emergency telephone system, the Voicenet system provides a means to communicate with other corporate locations with which the plant has a need to communicate. This system bypasses external commercial telephone lines and switching equipment. Corporate transmission facilities provide fiber optic, copper-wire, and microwave radio to ensure a high degree of system reliability. In addition to the redundancy provided by the three system options, backup power is provided for the systems. DEMNET is the primary means of offsite communication. This circuit allows intercommunication among the EOF, TSC, control room, counties, and states. DEMNET operates as an internet based (VoIP)communications system with a satellite back-up. Should the internet transfer rate become slow or unavailable, the DEMNET will automatically transfer to satellite mode.OPEP-02.2.1 Rev. 6 Page 122 of 310 ATTACHMENT 1 Page 97 of 219 EAL Bases Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by Duke Energy in four ways: (1) tie-ins through the PBX to any other plant location, (2) lines to plant emergency facilities, (3) lines to the Joint Information Center for public information purposes, and (4) lines to the AEF. The local service provider provides primary and secondary power for their lines at the Central Office.Satellite Phones A total of three portable satellite telephones are available which enable communication when all other phone systems are inoperable, e.g. following a major hurricane. These portable systems can be powered by internal batteries, external DC sources as well as external AC sources. Two of these phones require outside use, while one phone may used either outside or in the EOF with a permanently mounted external antenna.Cellular Phones Selected plant personnel are provided with cellular telephones. These phones may be used during emergencies if other communications means are not readily available or are inoperable. These phones are not expected to be used in the Control Room or Power Block due to interference with plant equipment and loss of signal to the phone.NRC Emergency Telecommunications System The NRC uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRC regional and national offices. The Duke Energy communications line provides a link independent of the local public telephone network.Telephones connected to this network are located in the Brunswick Control Room, Technical SuppoA Center, and Emergency Operations Facility. These lines will not function if the PBX Telephone System fails.This EAL is the cold condition equivalent of the hot condition EAL SU7.1.NEI 99-01 Basis: This IC addresses a significant loss of onrsite or offsite communications capabilities. While-not Deleted:-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 ornsite_ -{Deleted: -information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).IOPEP-02.2.1 ev. 6 Page 123 of 310 The first EAL condition addresses a total loss of the communications methods used in support Deleted: EAL#1 of routine plant operations. /.The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here aretheState .......Brunswick and New Hanover County EOCs The third EAL addresses a total loss of the communications methods used to notify the NRC -of an emergency declaration., J Deleted: EAL #2 Deleted: (see Developer Notes)" Deleted:.-Deleted: EAL #3 ATTACHMENT 1 Page 98 of 219 EAL Bases BNP Basis Reference(s):

1. OERP Radiological Emergency Response Plan Appendix A 2. SD-48 Communication Systems 3. NEI 99-01 CU5 IOPEP-02.2.1 I Rev. 6 Page 124 of 310 ATTACHMENT 1 Page 99 of 219 EAL Bases Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 6 -Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL: CA6.1 Alert The occurrence of any Table C-5 hazardous event AND 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 Table C-5 Hazardous Events* Seismic event (earthquake)
Internal or external FLOODING event* High winds or tornado strike* FIRE* EXPLOSION* Other events with similar hazard characteristics as determined by the Shift Manager Mode Applicability:

4 -Cold Shutdown, 5 -Refuel I OPEP-02.2.1 Rev. 6 1 Page 125 of 310 1 ATTACHMENT 1 Page 100 of 219 EAL Bases Definition(s): 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 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.FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.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 (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. 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. BNP Basis: " The significance of seismic events are discussed under EAL HU2.1 (ref. 1)." Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 2, 3)." Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 135 mph. (ref. 4).OPEP-02.2.1 IRev. 6 Page 126 of 310 ATTACHMENT 1 Page 101 of 219 EAL Bases* Areas containing functions and systems required for safe shutdown of the plant are identified by Fire Zone in the fire response procedure (ref. 5, 6).* An explosion that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL.NEI 99-01 Basis: This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode.This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant..The first conditional addresses damaage to a SAFETY SYSTEM train that is in -Deleted: EAL .1 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 conditional addresses damage to a SAFETY SYSTEM component that is not in --Deleted: EAL 1.b.2 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 CS1 or.RS1. ---------------- Deleted: AS1 BNP Basis Reference(s):

1. 1(2)APP-UA-28 6-4 Seismic Event 2. OAOP-13.0 Operations During Hurricane, Flood Conditions, Tornado or Earthquake

3. Updated FSAR section 3.4.2 Protection From Internal Flooding 4. Updated FSAR Section 2.3.1.2.7 5. BNP-E-9.004 Safe Shutdown Analysis Report 6. OPFP-PBAA Power Block Auxiliary Areas Prefire Plan 7. NEI 99-01 CA6 OPEP-02.2.1 ev. 61 Page 127 of 310 ATTACHMENT 1 Page 102 of 219 EAL Bases Category H -Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety.1. Security Unauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.3. Natural or Technology Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.

4. Fire Fires can pose significant hazards to personnel and reactor safety. Appropriate for classification are fires within the site Protected Area or which may affect operability of equipment needed for safe shutdown 5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant.OPEP-02.2.1 ev. 6 Page 128 of 310 ATTACHMENT 1 Page 103 of 219 EAL Bases 6. Control Room Evacuation Events that are indicative of loss of Control Room habitability.

If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.

7. SEC Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification.

While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Site Emergency Coordinator the latitude to classify emergency conditions consistent with the established classification criteria based upon Site Emergency Coordinator judgment.OPEP-02.2.1 Rev. 6 Page 129 of 310 ATTACHMENT 1 Page 104 of 219 EAL Bases Category: H -Hazards Subcategory: 1 -Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL: HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervision Mode Applicability: All Definition(s): 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.HOSTILE ACTION- An act toward BNP 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 BNP. 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).BNP Basis: The Security Shift Supervision is defined as either the Security Shift Lieutenant or the Security Shift Sergeant.This EAL is based on the BNP Physical Security Plan (ref. 1).I OPEP-02.2.1 Rev. 6 1 Page 130 of 310 ATTACHMENT 1 Page 105 of 219 EAL Bases NEI 99-01 Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HG1.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 Offsite Response Organizations. 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 Plant ------.-- ---Deleted: [and Independent Spent Fuel Storage Installation Security Program].his EAL references the.Security Shift Supervison because these are the individuals trained to -Deleted: EAL #1 confirm that a security event is occurring or has occurred. Training on security event .Deleted: (site-specific confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § ' Deleted: security 2.39 information. Deleted: shift supervision) Emeqrgerncyplans and implemrenting 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 BNP Physical Security Plan (ref. 1).Escalation of the emergency classification level would be via IC HAl.Deleted: EAL #2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with (site-specific procedure).¶ EAL #3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.The status and size of the plane may also be provided by NORAD through the NRC.Validation of the threat is performed in accordance with (site-specific procedure).¶ I OPEP-02.2.1 I Rev. 6 Page 131 of 310 ATTACHMENT 1 Page 106 of 219 EAL Bases BNP Basis Reference(s):

1. BNP Physical Security Plan 2. SEC-NGGC-2170 Security Event Procedures

3. OAOP-40.0 Security Events 4. NEI 99-01 HU1 I OPEP-02.2.1 I Rev. 6 Page 132 of 310 ATTACHMENT 1 Page 107 of 219 EAL Bases Category: Subcategory:

H -Hazards 1 -Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL: HU1.2 Unusual Event Notification of a credible security threat directed at the site Mode Applicability: All Definition(s): None BNP Basis: The Security Shift Supervision is defined as either the Security Shift Lieutenant or the Security Shift Sergeant.This EAL is based on the BNP Physical Security Plan (ref. 1).NEI 99-01 Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HG1.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 Offsite Response Organizations. 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 Plar.A:7ACHMENT 1 Page 107 of 219 EAL Bases OPEP-02.2.1 Rev. 6 Page 133 of 310 Deleted: [and Independent Spent Fuel Storage Installation Security Program]I Deleted: EAL #1 references (site-specific security shift supervision) because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information.1 This EAL addresses the receipt of a credible security threat. The credibility of the threat is , -j Deleted: EAL #2 assessed in accordance

t-si-a--e-----t-

---- ------ ----- --------------

-4 Deleted: (site-specific procedure) Deleted: EAL #3 addresses the threat from the ,_Emferge nrcy_pjans and implementing procedures are public documents; therefore, EALs should impact of an aircraft on the plant. The NRC not incorporate Security-sensitive information. This includes information that may be 1 Headquarters Operations Officer (HOO) will communicate to the licensee if the threat advantageous to a potential adversary, such as the particulars concerning a specific threat or involves an aircraft. The status and size of the threat location. Security-sensitive information should be contained in non-public documents plane may also be provided by NORAD through Ssuch as the BNP Physical Security Plan (ref. 1'). the NRC. Validation of the threat is performed in accordance with (site-specific procedure).¶I Escalation of the emergency classification level would be via IC HAl.BNP Basis Reference(s):

1. BNP Physical Security Plan 2. SEC-NGGC-2170 Security Event Procedures

3. OAOP-40.0 Security Events 4. NEI 99-01 HU1 IOPEP-02.2.1 I Rev. 6 1 Page 134 of 310 ATTACHMENT 1 Page 109 of 219 EAL Bases Category: Subcategory:

H -Hazards 1 -Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL: HU1.3 Unusual Event A validated notification from the NRC providing information of an aircraft threat Mode Applicability: All Definition(s): None BNP Basis: The Security Shift Supervision is defined as either the Security Shift Lieutenant or the Security Shift Sergeant.This EAL is based on the BNP Physical Security Plan (ref. 1).NEI 99-01 Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HG1.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 Offsite Response Or-ganizations. Security plans and terminology are based on the guidance provided by NEI 03-12, Template -Deleted: [and Independent Spent Fuel for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan .... ---. Storage Installation Security Program]_ACHMENT 1 Page 109 of 219 EAL Bases Deleted: EAL #1 references (site-specific 1 security shift supervision) because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information.¶ I OPEP-02.2.1 Rev. 6 1 Page 135 of 310 .This EAL addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with the BNP Physical Security Plan (ref. .1t 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 BNP Physical Security Plan (ref. 1).Escalation of the emergency classification level would be via IC HAl.--Deleted: EAL #2 addresses the receipt of a 1 credible security threat. The credibility of the threat is assessed in accordance with (site-specilic procedure).¶ EAL #3--J Deleted: (site-specific procedure) I BNP Basis Reference(s):

1. BNP Physical Security Plan 2. SEC-NGGC-2170 Security Event Procedures

3. 0AOP-40.0 Security Events 4. NEI 99-01 HU1 IOPEP-02.2.1 I Rev. 6 1 Page 136 of 310 ATTACHMENT 1 Page 111 of 219 EAL Bases Category: Subcategory:

H -Hazards 1 -Security Initiating Condition: Hostile action within the owner controlled area or airborne attack threat within 30 minutes EAL: HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervision Mode Applicability: All Definition(s): HOSTILE ACTION- An act toward BNP 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 BNP. 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 -Area depicted as the property boundary in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan.BNP Basis: The Security Shift Supervision is defined as either the Security Shift Lieutenant or the Security Shift Sergeant.IOPEP-02.2.1 I Rev. 6 1 Page 137 of 310 ATTACHMENT 1 Page 112 of 219 EAL Bases NEI 99-01 Basis: This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.Timely and accurate communications between the 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 Pla .---------- --Deleted: [and Independent Spent Fuel I 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 Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions.This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72..This EAL is applicable for any HOSTILE ACTION occurrinq, or that has occurred, in-the ------OWNER CONTROLLED AREA.------------------------------------j Deleted: EAL #1--I Deleted: This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA.Deleted: EAL #2 addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes.The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with (site-specific procedure).¶ 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.¶IOPEP-02.2.1 I Rev. 6 Page 138 of 310 ATTACHMENT 1 Page 113 of 219 EAL Bases Ernmergerncyplans and implementing procedures are puyblicd 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 the1BNP Physical Security Plan (ref. 1).BNP Basis Reference(s):

1. BNP Physical Security Plan 2. SEC-NGGC-2170 Security Event Procedures

3. OAOP-40.0 Security Events 4. NEI 99-01 HA1 Deleted: 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.¶I Deleted: Security I OPEP-02.2.1 Rev. 6 1 Page 139 of 3101 ATTACHMENT 1 Page 114 of 219 EAL Bases Category: Subcategory:

H -Hazards 1 -Security Initiating Condition: Hostile action within the owner controlled area or airborne attack threat within 30 minutes EAL: HA1.2 Alert A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability: All Definition(s): None BNP Basis: The Security Shift Supervision is defined as either the Security Shift Lieutenant or the Security Shift Sergeant.NEI 99-01 Basis: This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.Timely and accurate communications between the 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 ........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 (OROs), 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 OPEP-02.2.1 IRev. 6 1Page 140 of 310--Deleted: [and Independent Spent Fuel Storage Installation Security Program]I 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..This EAL addresses the threat from the impact of an aircraft on the plant, and the anticipatedA arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance withsite-specific security procedures-,-Deleted: EAL #1 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 #2 SDeleted: Deleted: )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 BNP Physical Security Plan_ ................................. BNP Basis Reference(s):

1. BNP Physical Security Plan 2. SEC-NGGC-2170 Security Event Procedures

3. OAOP-40.0 Security Events 4. NEI 99-01 HA1__ -1 Deleted: Security I OPEP-02.2.1 I Rev. 6 Page 141 of 310 ATTACHMENT 1 Page 116 of 219 EAL Bases Category: Subcategory:

H -Hazards 1 -Security Initiating Condition: Hostile Action within the Protected Area EAL: HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision Mode Applicability: All Definition(s): HOSTILE ACTION- An act toward BNP 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 BNP. 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 double-fenced security area with intrusion detection devices immediately surrounding the plant structures. The Protected Area is depicted in BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations. BNP Basis: The Security Shift Supervision is defined as either the Security Shift Lieutenant or the Security Shift Sergeant.These individuals are the designated orqsite personnel/qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the BNP Physical Security Plan (Safeguards) information. (ref. 1)-4 Deleted: -I OPEP-02.2.1 I Rev. 6 1 Page 142 of 310 ATTACHMENT 1 Page 117 of 219 EAL Bases NEI 99-01 Basis: This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA.This event will require rapid response and assistance due to the possibility for damage to plant equipment. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plar --.----- ---Deleted: [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 Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions..This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.Emergency plans and implementing procedures are public documents; therefore, EALs 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 BNP Physical SecurityP/an_(re.1. ................................. -Deleted: 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 HAl. It also--Deleted: Safeguards Escalation of the emergency classification level would be via IC HG1.BNP Basis Reference(s):

1. BNP Physical Security Plan 2. SEC-NGGC-2170 Security Event Procedures

3. OAOP-40.0 Security Events 4. NEI 99-01 HS1 oPEP-02.2.1 Rev. 6 Page 143 of 310 Category: Subcategory:

Initiating Condition: ATTACHMENT 1 Page 118 of 219 EAL Bases H -Hazards 1 -Security Hostile Action resulting in loss of physical control of the facility EAL: HG1.1 General Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision AND EITHER of the following has occurred: Any of the following safety functions cannot be controlled or maintained

Reactivity" RPV water level" RCS heat removal OR Parnage to spent fuel has occurred or is IMMINENT-4 Formatted:

Font: Bold I Mode Applicability: All Definition(s): HOSTILE ACTION -An act toward BNP 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 BNP. 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).OPEP-02.2.1 Rev. 6 Page 144 of 310 ATTACHMENT 1 Page 119 of 219 EAL Bases 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 double-fenced security area with intrusion detection devices immediately surrounding the plant structures. The Protected Area is depicted in BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations. BNP Basis: The Security Shift Supervision is defined as either the Security Shift Lieutenant or the Security Shift Sergeant.NEI 99-01 Basis: This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions. It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan ---------- -Deleted: [and Independent Spent Fuel I Storage Installation Security Program]Emergency plans and implementing procedures are public documents; therefore, EALs should Deleted: 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 BNP Physical Security Plan (ref.1).BNP Basis Reference(s):

1. BNP Physical Security Plan 2. SEC-NGGC-2170 Security Event Procedures

3. OAOP-40.0 Security Events 4. NEI 99-01 HG1 IOPEP-02.2.1 IRev. 6 1Page 145 of 310 ATTACHMENT 1 Page 120 of 219 EAL Bases Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 -Seismic Event Initiating Condition: Seismic event greater than OBE levels EAL: HU2.1 Unusual Event Seismic event > OBE per OAOP-1 3.0 Mode Applicability: All Definition(s): None BNP Basis: Ground motion acceleration of 0.08g is the Operating Basis Earthquake for BNP (ref. 1).Unit 2 has an active Kinemetrics Condor Seismic Monitoring System with the following components used for seismic detection for the Brunswick Site: The system will detect and digitally record the response to actual earthquake loading in terms of acceleration time history from the existing accelerometers mounted in the Unit 2 -17ft. elevation (basement) of the Reactor Building and also at +89 foot elevation mounted on the Reactor Containment structure. The system will automatically evaluate the recorded acceleration time history in order to determine the response spectra of the events and compare those to the Operating Basis Earthquake (OBE) parameters graphically. It will also determine the exceedance of the OBE, and provides a hard copy of this comparison. The system will provide an immediate Event Alarm output signal at a trigger threshold value of 0.01 g to alarm the existing Annunciator 1 (2)UA-28 6-4 SEISMIC EVENT in the Control Room back to alert the Operators to a seismic event. (ref. 1, 2)The BNP seismic instrumentation supports readily assessable OBE indications (> 0.08g acceleration) within the Control Room at panel 2-ENV-XU-823. OAOP-1 3.0 provides the guidance for determining if the OBE earthquake threshold is exceeded. (ref. 3).The Shift Manager or Site Emergency Coordinator 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. 0PEP-02.2.1 IRev. 6 IPage 146 of 310 ATTACHMENT 1 Page 121 of 219 EAL Bases To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center) can confirm that an earthquake has occurred in the area of the plant. As stated above, such confirmation should not, however, preclude a timely emergency declaration. The NEIC can be contacted by calling (303) 273-8500. Select option #1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of BNP. Provide the analyst with the following BNP coordinates: 330 57' 30" north latitude, 780 00' 30" west longitude (ref. 4). Alternatively, near real-time seismic activity can be accessed via the NEIC website: http://earthquake.usgs.gov/eqcenterl NEI 99-01 Basis: This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.Event verification with external sources should not be necessary during or following an OBE.Earthquakes of this magnitude should be readily felt by or~ite personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.08g). The Shift Manager or Site Emergency Coordinator 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 -Deleted: -tDjd.ted. Director Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 orA.8 ........................... j -Deleted: SA9 I OPEP-02.2.1 I Rev. 6 1 Page 147 of 310 ATTACHMENT 1 Page 122 of 219 EAL Bases BNP Basis Reference(s):

1. Updated FSAR section 2.5.2.6 2. 1(2)APP-UA-28 6-4 Seismic Event 3. OAOP-13.0 Operations During Hurricane, Flood Conditions, Tornado or Earthquake

4. Updated FSAR section 2.1.1.1 5. NEI 99-01 HU2 OPEP-02.2.1 Rev. 6 Page 148 of 310 ATTACHMENT 1 Page 123 of 219 EAL Bases Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability: All Definition(s): PROTECTED AREA -The double-fenced security area with intrusion detection devices immediately surrounding the plant structures. The Protected Area is depicted in BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations. Deleted: EA Deleted: EA BNP Basis: building room BNP Bais: ,isolating pow component d if concerns. C If damage is confirmed visually or by other in-plant indications, the event may be escalated to water level o an Alert under EAL CA6.1 or SA8.1. automatic isc fI component fi If breaker or re A tornado striking (touching down) within the Protected Area warrants declaration of an 11 operability of Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado I required by T current opera is defined as a violently rotating column of air in contact with the ground and extending from EAL #3 addr, the base of a thunderstorm. originating at magnitude to I personnel wil NEI 99-01 Basis: If EAL #4 addr ,i causes an or movement ai This IC addresses hazardous events that are considered to represent a potential degradation the plant staf of the level of safety of the plant. personal veh I include site fl r heavy rains, ,EAL HU3.1 addresses a tornado striking (touching _down) within the PROTECTED AREA. failure , etc., blocking the This EAL is ,Escalation of the emergency classification level would be based on ICsin Recognition impediments a r F S o C. ----------- breakdowns Categories S F or -C.---------------------------------------------------- significant co\ Andrew strik*BNP Basis Reference(s): ' flooding arou Midwest flooq 1. NEI 99-01 HU3 Ft. Calhoun EAL #5 addr I Deleted: A ,L #1 L #2 addresses flooding of a or area that results in operators rer to a SAFETY SYSTEM ue to water level or other wetting lassification is also required if the r related wetting causes an olation of a SAFETY SYSTEM rom its power source (e.g., a lay trip). To warrant classification, the affected component must be echnical Specifications for the Bting mode. 1"sses a hazardous materials event an offsite location and of sufficient impede the movement of thin the PROTECTED AREA.1 sses a hazardous event that n-site impediment to vehicle nd significant enough to prohibit If from accessing the site using icles. Examples of such an event ooding caused by a hurricane, up-river water releases, dam or an on-site train derailment access road. ¶not intended apply to routine such as fog, snow, ice, or vehicle or accidents, but rather to more inditions such as the Hurricane e on Turkey Point in 1992, the nd the Cooper Station during the ds of 1993, or the flooding around Station in 2011 .¶eassea (site-specific description). I OPEP-02.2.1 Rev. 6 1 Page 149 of 310 ATTACHMENT 1 Page 124 of 219 EAL Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.2 Unusual Event 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 Mode Applicability: All Definition(s): FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.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 (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. BNP Basis: Refer to Updated FSAR section 3.4.2 Protection From Internal Flooding to identify susceptible internal flooding areas (ref. 1).1OPEP-02.2.1 I Rev. 6 1 Page 150 of 310 ATTACHMENT 1 Page 125 of 219 EAL Bases NEI 99-01 Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.IThis EAL 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.=,Escalation of the emerg ency classification level would be based on ICsin Recognition Cateaories_, F, SorC.BNP Basis Reference(s):

1. Updated FSAR section 3.4.2 Protection From Internal Flooding 2. NEI 99-01 HU3 Deleted: EAL #1 addresses a tornado striking (touching down) within the PROTECTED AREA.¶Deleted: EAL #3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.¶EAL #4 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles.

Examples of such an event include site flooding caused by a hurricane. heavy rains, up-river water releases, dam failure, etc.. or an on-site train derailment blocking the access road. ¶This EAL is not intended apply to 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.T EAL #5 addresses (site-specific description). Deleted: A I OPEP-02.2.1 Rev. 6 1 Page 151 of 3101 ATTACHMENT 1 Page 126 of 219 EAL Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.3 Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)Mode Applicability: All Definition(s): IMPEDE(D)- Personnel access to a room or area is hindered to an extent that 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). PROTECTED AREA -The double-fenced security area with intrusion detection devices immediately surrounding the plant structures. The Protected Area is depicted in BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations. BNP Basis: The following documents provide additional information on hazardous substances and spills." OAOP-34.0 Chlorine Emergencies (Ref. 1)* OAOP-44.0 Sodium Hypochlorite or Acti-Brom Leak Response (Ref. 2)* OAOP-43.0 Hydrogen Emergency (Ref. 3)* OAOP-05.0 Radioactive Spills, High Radiation, and Airborne Activity (Ref. 4)" Regulatory Guide 1.78, Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release, Table 1, Toxicity Limits (IDLH Limits) for Some Hazardous Chemicals (Ref. 5)OPEP-02.2.1 Rev. 6 Page 152 of 310 ATTACHMENT 1 Page 127 of 219 EAL Bases NEI 99-01 Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.This EAL addresses a hazardous materials event originating -t t DDeleted: EAL #1 addresses a tornado striking hs A a re e-rnltothePROTECT __ _ -(touching down) within the PROTECTED AREA and of sufficient magnitude to impede the movement of personnel within the AREA.¶This EAL addresses flooding of a building room PROTECTED AREA. I or area that results in operators isolating power to a SAFETY SYSTEM component due to water ,Escalation of the emergency classification level lwould be based on ICs inRecognition , evelor other wetting concerns. Classification -is also required if the water level or related CategoriesB, F, S or C. -t wetting causes an automatic isolation of a SAFETY SYSTEM component from its power B ' source (e.g., a breaker or relay trip). To warrant BNP Basis Reference(s):' classification, operability of the affected i component must be required by Technical 1. OAOP-34.0 Chlorine Emergencies %x 1, Specifications for the current operating mode.,\ ' EAL#3 2. OAOP-44.0 Sodium Hypochlorite or Acti-Brom Leak Response ' el at\1Deleted: at an offsite location 3. OAOP-43.0 Hydrogen Emergency Deleted: EAL #4 addresses a hazardous event 4. 0AOP-05.0 Radioactive Spills, High Radiation, and Airborne Activity that causes an on-site impediment to vehicle I movement and significant enough to prohibit 5. Regulatory Guide 1 .78, Evaluating the Habitability of a Nuclear Power Plant Control Room the plant staff from accessing the site using personal vehicles. Examples of such an event During a Postulated Hazardous Chemical Release, Table 1, Toxicity Limits (IDLH Limits) include site flooding caused by a hurricane, for Some Hazardous Chemicals heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment

6. NEI 99-01 HU3 blocking the access road. T This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more ATTACHMENT 1 significant conditions such as the Hurricane Page 127 of 219 Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the EAL Bases Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 201 1¶EAL #5 addresses (site-specific description).

Category: H -Hazards and Other Conditions Affecting Plant Safety Deleted: A Subcategory: 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.4 Unusual Event A hazardous event that results in onsite conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7)Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. Mode Applicability: I OPEP-02.2.1 Rev. 6 1 Page 153 of 310 All Definition(s): None BNP Basis: The 15 minute clocks starts when the ORO Director of Emergency Services (Brunswick and New Hanover Counties) and the Shift Manager agree that Onsite/Offsite conditions are sufficient to prohibit the plant staff from accessing the site via personal vehicles.NEI 99-01 Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.-------........................................................... Deleted: EAL #1 addresses a tornado striking (touching down) within the PROTECTED AREA.¶This EAL addresses flooding of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.¶EAL #3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.¶I OPEP-02.2.1 Rev. 6 1 Page 154 of 310 ATTACHMENT 1 Page 129 of 219 EAL Bases , d Deleted: EAL #4 This EAL addresses a hazardous event, either onsite or offsite, that causes an onsite 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 onsite 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._ __- -Deleted:-j Escalation of the emergency classification level would be based on ICs in Recognition Categories P, _F, S or_C -.-........

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

BNP Basis Reference(s):

1. NEI 99-01 HU3-Deleted: EAL #5 addresses (site-specific

-- description). I j -Deleted: A I I OPEP-02.2.1 Rev. 6 1 Page 155 of 310 ATTACHMENT 1 Page 130 of 219 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety Category: Subcategory: 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.5 Unusual Event Intake Canal water level > +19 ft Mean Sea Level OR Intake Canal water level < -7.75 ft Mean Sea Level Mode Applicability: All BNP Basis: The high Intake Canal level is the highest remotely measurable Intake Canal water level.Otherwise it would have been based based the plant design that Class I structures and engineered safety features systems are protected against still water flooding (elevation 22.0 feet). BNP is geographically located in close proximity to the Atlantic coastal storm track and has an approximate grade elevation of 20 feet above Mean Sea Level. Hurricanes and tropical storms are therefore, the most extreme weather phenomena that affect the site area. Potential subsequent flooding should be considered even though the plant structures were designed to compensate, via installed sump pumps, for a maximum site flooding depth of 22 feet above Mean Sea Level during the Maximum Probable Hurricane. (ref. 1).The minimum water level predicted for the Maximum Probable Hurricane is -7.5 feet Mean Sea Level under special case circumstances. The abnormal operating procedure for a hurricane requires that each unit be shutdown prior to arrival of hurricane conditions at the site. The SW System has been analyzed in modes 4 and 5 for an intake canal water level of-7.75 feet Mean Sea Level corresponding to -8.63 feet Mean Sea Level in the pump suction bay for the maximum pressure drop, 0.88 feet, across the traveling screens. (ref. 2, 3).IOPEP-02.2.1 I Rev. 6 Page 156 of 310 ATTACHMENT 1 Page 131 of 219 EAL Bases NEI 99-01 Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.This EAL addresses high and low external water levels as a result of a hurricane. Escalation of the emergency classification level would be based on ICs in Recognition Categories, F, S or C.BNP Basis Reference(s):

1. Updated FSAR section 2.4.10.2 2. Updated FSAR section 9.2.1.2.3 3. OAOP-13.0 Operations During Hurricane, Flood Conditions, Tornado or Earthquake II II II II II II II II II II II Deleted: EAL #1 addresses a tornado striking (touching down) within the Protected Area.¶EAL #2 addresses flooding of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.

Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.EAL #3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.¶EAL #4 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff 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 #5 Deleted: (site-specific description) Deleted: A tl IOPEP-02.2.1 I Rev. 6 Page 157 of 310 ATTACHMENT 1 Page 132 of 219 EAL Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 4- Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1): " 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 The FIRE is located within any Table H-1 area Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table H-1 Fire Areas" Reactor Building* Diesel Generator Building* Diesel 4-Day Tank Rooms" Service Water Building" Turbine Building" Control Building" CSTs* Diesel Fuel Oil Storage Tank Mode Applicability:

All I OPEP-02.2.1 I Rev. 6 Page 158 of 310 ATTACHMENT 1 Page 133 of 219 EAL Bases Definition(s): 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.BNP Basis: The 15 minute requirement begins with a credible notification that a fire is occurring, or receipt of multiple valid fire detection system alarms or field validation of a single fire alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 15 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, the 15 minute time limit is from the original receipt of the fire detection alarm.Table H-1 Fire Areas are based on BNP-E-9.004 Safe Shutdown Analysis Report and OPFP-PBAA Power Block Auxiliary Areas Prefire Plan. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1, 2).NEI 99-01 Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.For EAL HU4.1 the intent of the 15-minute duration is to size the FIRE and to discriminate -against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or .Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8.BNP Basis Reference(s):

1. BNP-E-9.004 Safe Shutdown Analysis Report 2. OPFP-PBAA Power Block Auxiliary Areas Prefire Plan 3. NEI 99-01 HU4 OPEP-02.2.1 Rev. 6 Page 159 of 310 Deleted: EAL #1 1 The Deleted: EAL #2 T This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.¶ A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists;however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.¶ If an actual FIRE is verified by a report from the field, then EAL #1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.

¶EAL #3 T1 In addition to a FIRE addressed by EAL #1 or EAL #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.[Sentence for plants with an ISFSI outside the plant Protected Areas¶EAL #4 ¶If a FIRE within the plant or ISFSI [for plants with an ISFS1 outside the plant Protected Area]PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.Deleted: SA9 ATTACHMENT 1 Page 134 of 219 EAL Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 -Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE)AND The fire alarm is indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table H-1 Fire Areas* Reactor Building* Diesel Generator Building* Diesel 4-Day Tank Rooms" Service Water Building* Turbine Building" Control Building" CSTs" Diesel Fuel Oil Storage Tank Mode Applicability: All IOPEP-02.2.1 I Rev. 6 1 Page 160 of 310 ATTACHMENT 1 Page 135 of 219 EAL Bases Definition(s): 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.BNP Basis: The 30 minute requirement begins upon receipt of a single valid fire detection system alarm.The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1.Table H-1 Fire Areas are based on BNP-E-9.004 Safe Shutdown Analysis Report and OPFP-PBAA Power Block Auxiliary Areas Prefire Plan. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1, 2).NEI 99-01 Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.I ~Deleted: EAL #1 T]IThis EAL addresses receipt of a s firealarm, and the existence of a FIRE is not verified -The intent of the 15-minute duration is to size (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, smoldering waste paper basket). In addition to the 30-minute clock starts at the time that the initial alarm was received, and not the time that alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation asubsequent verification action was performed. of a suppression system, etc. t Upon receipt, operators will take prompt actions ATTACHMENT 1 to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment Page 135 of 219 purposes, the emergency declaration clock starts at the time that the initial alarm, EAL Bases 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 A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure alarm, indication or report.T or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to EAL#2 T 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 U4.1 is im e ate applicable, -Deleted: EAL #and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and I OPEP-02.2.1 I Rev. 6 Page 161 of 310 this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. pepending upon the plant mode at the time of the event,eescalation of the emergency.. classification level would be via IC CA6 or SA8.BNP Basis Reference(s):

1. BNP-E-9.004 Safe Shutdown Analysis Report 2. OPFP-PBAA Power Block Auxiliary Areas Prefire Plan 3. NEI 99-01 HU4 Deleted: EAL #3 T In addition to a FIRE addressed by EAL #1 or EAL #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.[Sentence for plants with an ISFSI outside the plant Protected Area]¶EAL #4 ¶Deleted: SA9 I OPEP-02.2.1 I Rev. 6 1 Page 162 of 310o ATTACHMENT 1 Page 137 of 219 EAL Bases Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 -Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.3 Unusual Event A FIRE within the plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: All Definition(s): 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.PROTECTED AREA -The double-fenced security area with intrusion detection devices immediately surrounding the plant structures. The Protected Area is depicted in BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations. BNP Basis: None ,, NEI 99-01 Basis: ,i 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. h,, ,'I Jn addition to a FIRE addressed by EAL.HU4.1 orUU4.2, a FIRE within theplant -, , PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level ','of plant safety. ,-------------------------------------------------------- j OPEP-02.2.1 Rev. 6 I Page 163 of 3101 Deleted: EAL #1 T The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system. etc. ¶`Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was pertormed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.¶EAL #2 ¶This EAL addresses receipt ot a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.¶ A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists;however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.¶ If an actual FIRE is verified by a report from the field, then EAL #1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. ¶EAL #3 ¶Deleted: #1 Deleted: EAL #2 Deleted: This basis extends to a FIRE occurring within the PROTECTED AREA of an/SFSI located outside the plant PROTECTED AREA. [Sentence for plants with an ISFSI outside the plant Protected Area]Deleted: EAL #4 '[If a FIRE within the plant or ISFSI [forplants with an ISFSI outside the plant Protected Area]PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary F-Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8 ---------------------------------- BNP Basis Reference(s):

1. NEI 99-01 HU4--Deleted: 9 IOPEP-02.2.1 IRev. 6 1Page 164 of 310 ATTACHMENT 1 Page 139 of 219 EAL Bases Category: Subcategory:

H -Hazards and Other Conditions Affecting Plant Safety 4 -Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.4 Unusual Event A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability: All Definition(s): 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.PROTECTED AREA -The double-fenced security area with intrusion detection devices immediately surrounding the plant structures. The Protected Area is depicted in BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations. BNP Basis: 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. I OPEP-02.2.1 Rev. 6 1 Page 165 of 310 ATTACHMENT 1 Page 140 of 219 EAL Bases NEI 99-01 Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.Jfa FIRE within the plantpROTECTED AREA is of sufficient size to_-e-qirearesp__ nse 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.Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA.BNP Basis Reference(s):

1. NEI 99-01 HU4~1~1I II II II II II II II II II II II II II II II II II Deleted: EAL #1 T1 The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldedrng waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. ¶Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed.

Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.¶EAL #2 ¶This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.¶ A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists: however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.¶ If an actual FIRE is verified by a report from the field, then EAL #1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. T]EAL #39¶In addition to a FIRE addressed by EAL #1 or EAL #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.[Sentence for plants with an ISFSI outside the plant Protected Area]¶I EAL #4 ¶Deleted: or ISFSI [for plants with an ISFSI outside the plant Protected Area]Deleted: 9 IOPEP-02.2.1 I Rev. 6 Page 166 of 310 ATTACHMENT 1 Page 141 of 219 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety Category: Subcategory: 5 -Hazardous Gases Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL: HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 rooms or areas AND Entry into the room or area is prohibited or IMPEDED (Note 5)Note 5: 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. Table H-2 Safe Operation & Shutdown Areas Room/Area Mode Applicability Reactor Building -17' North RHR Unit-1 & 2 3, 4, 5 Reactor Building -17' South RHR Unit-1 & 2 3, 4, 5 Reactor Building 20' East & West MCC Areas Unit-1 & 2 3, 4, 5 Reactor Building 20' Pipe Tunnel Unit-1 & 2 3, 4, 5 Mode Applicability: All Definition(s): IMPEDE(D)- Personnel access to a room or area is hindered to an extent that 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). BNP Basis: If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.OPEP-02.2.1 Rev. 6 Page 167 of 310 ATTACHMENT 1 Page 142 of 219 EAL Bases The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).The following documents provide additional information on hazardous substances and spills.* OAOP-34.0 Chlorine Emergencies (Ref. 2)" OAOP-44.0 Sodium Hypochlorite or Acti-Brom Leak Response (Ref. 3)* OAOP-43.0 Hydrogen Emergency (Ref. 4)NEI 99-01 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 ..te -- -Deleted: Emergency Director Emergency Coordinator'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). IOPEP-02.2.1 Rev. 6 Page 168 of 310o ATTACHMENT 1 Page 143 of 219 EAL Bases An emergency declaration is not warranted if any of the following conditions apply: " The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release). For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4." The gas release is a 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." If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.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

-Deleted: (BWR only)Escalation of the emergency classification level would be via Recognition CategoryR C or F Deleted: A ICs.BNP Basis Reference(s):

1. Attachment 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases 2. OAOP-34.0 Chlorine Emergencies

3. OAOP-44.0 Sodium Hypochlorite or Acti-Brom Leak Response 4. OAOP-43.0 Hydrogen Emergency 5. NEI 99-01 HA5 I OPEP-02.2.1 Rev. 6 1 Page 169 of 310 ATTACHMENT 1 Page 144 of 219 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety Category: Subcategory:

6 -Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL: HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panels Mode Applicability: All Definition(s): None BNP Basis: The Shift Manager (SM) determines if the Control Room is inoperable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions (Ref. 1, 2).Inability to establish plant control from outside the Control Room escalates this event to a Site Area Emergency per EAL HS6.1.NEI 99-01 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. OPEP-02.2.1 I Rev. 6 Page 170 of 310 ATTACHMENT 1 Page 145 of 219 EAL Bases Escalation of the emergency classification level would be via IC HS6.BNP Basis Reference(s):

1. OAOP-32.0, Plant Shutdown from Outside Control Room 2. OPLP-011.5 Alternate Shutdown Capability Controls 3. NEI 99-01 HA6 I OPEP-02.2.1 I Rev. 6 Page 171 of 310 ATTACHMENT 1 Page 146 of 219 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 6 -Control Room Evacuation Category: Subcategory:

Initiating Condition: Inability to control a key safety function from outside the Control Room EAL: HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panels AND Control of any of the following key safety functions is not reestablished within 22.5 min.(Note 1): " Reactivity" RPV water level* RPV pressure Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: All Definition(s): None BNP Basis: The Shift Manager determines if the Control Room is inoperable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions (Ref. 1, 2).The immediate actions of OAOP-32 direct a reactor scram prior to evacuating the Control Room thus no further action is required for reactivity control. Local control of high pressure injection sources and Safety Relief Valves (SRVs) establishes control of RPV water level and pressure.IOPEP-02.2.1 I Rev. 6 Page 172 of 310 ATTACHMENT 1 Page 147 of 219 EAL Bases NEI 99-01 Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Site EmergencyCoordinator Judgment. The Site Emergency ......Coordinator is expected to make -a reasonable, informed j ud-gm_ ent within_22.5 minutes _.whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s) (ref. 3).Escalation of the emergency classification level would be via IC FG1 or CG1 BNP Basis Reference(s):

1. OAOP-32.0, Plant Shutdown from Outside Control Room 2. OPLP-01.5 Alternate Shutdown Capability Controls 3. Calculation No. BNP-E-9.007 ASSD Manual Action Feasibility

4. NEI 99-01 HS6 I Deleted:iector

-ýDe11. iector~I Deleted: (the site-specific time for transfer) I OPEP-02.2.1 Rev. 6 Page 173 of 310 ATTACHMENT 1 Page 148 of 219 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety Category: Subcategory: 7 -SEC Judgment Initiating Condition: Other conditions existing that in the judgment of the Site Emergency Coordinator warrant declaration of a UE EAL: HU7.1 Unusual Event Other conditions exist which in the judgment of the Site Emergency Coordinator 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.--{ Formatted: Space Before: 6 pt I Mode Applicability: All Definition(s): None BNP Basis: The Site Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the BNP Emergency Response Plan. The Operations Shift Manager (SM) initially acts in the capacity of the Site Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Site Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).NEI 99-01 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 Site EmergencyCoordinator to fall under the eme[g!ency classification level description for an --_ --- Deleted: Director ,Unusual Event. .-Deleted: NOUE OPEP-02.2.1 Rev. 6 Page 174 of 310 ATTACHMENT 1 Page 149 of 219 EAL Bases BNP Basis Reference(s):

1. OERP BNP Radiological Emergency Response Plan section 3.0 Emergency Response Organization

2. NEI 99-01 HU7 I PEP-02.2.1 IRev. 6 1Page 175 of 310 ATTACHMENT 1 Page 150 of 219 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 7 -SEC Judgment Category: Subcategory:

Initiating Condition: Other conditions exist that in the judgment of the Site Emergency Coordinator warrant declaration of an Alert EAL: HA7.1 Alert Other conditions exist which, in the judgment of the Site Emergency Coordinator, 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.Mode Applicability: All Definition(s): HOSTILE ACTION- An act toward BNP 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 BNP. 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).I OPEP-02.2.1 I Rev. 6 Page 176 of 310 ATTACHMENT 1 Page 151 of 219 EAL Bases BNP Basis: The Site Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the BNP Emergency Response Plan. The Operations Shift Manager (SM) initially acts in the capacity of the Site Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Site Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref.1).NEI 99-01 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 Site EmergencyCoordinator to fall under the emergency classification level description for an Alert. --( Deleted: Director BNP Basis Reference(s):

1. OERP BNP Radiological Emergency Response Plan section 3.0 Emergency Response Organization

2. NEI 99-01 HA7 I I OPEP-02.2.1 Rev. 6 1 Page 177 of 310 ATTACHMENT 1 Page 152 of 219 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety Category: Subcategory:

7 -SEC Judgment Initiating Condition: Other conditions existing that in the judgment of the Site Emergency Coordinator warrant declaration of a Site Area Emergency EAL: HS7.1 Site Area Emergency Other conditions exist which in the judgment of the Site Emergency Coordinator 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 Mode Applicability: All Definition(s): .- -Formatted: Line spacing: single HOSTILE ACTION- An act toward BNP 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 BNP. 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)BNP Basis: The Site Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the BNP Emergency Response Plan. The Operations Shift Manager (SM) initially acts in the capacity of the Site Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Site Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation oPEP-02.2.1 Rev. 6 1 Page 178 of 310--Formatted: Space Before: 6 pt, After: 6 pt----{ Formatted: Space Before: 6 pt] of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).NEI 99-01 Basis:.- --Formatted: Space Before: 6 pt, After: 6 pt )This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Site EmergencyCoordinator to fall under the emergency classification level description for a Site Area Emergency. _ -J Deleted: Director BNP Basis Reference(s):

1. OERP BNP Radiological Emergency Response Plan section 3.0 Emergency Response Organization

2. NEI 99-01 HS7 OPEP-02.2.1 Rev. 6 Page 179 of 310 ATTACHMENT 1 Page 154 of 219 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 7 -SEC Judgment Category: Subcategory:

Initiating Condition: Other conditions exist which in the judgment of the Site Emergency Coordinator warrant declaration of a General Emergency EAL: HG7.1 General Emergency Other conditions exist which in the judgment of the Site Emergency Coordinator 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 Mode Applicability: All Definition(s): HOSTILE ACTION- An act toward BNP 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 BNP. 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.BNP Basis: The Site Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the BNP Emergency Response Plan. The Operations Shift Manager(SM) initially acts in the capacity of the Site Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Site Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is OPEP-02.2.1 Rev. 6 Page 180 of 310 responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref.1).Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the Site Boundary.NEI 99-01 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 Site EmergencyCoordinator to fall under the emergency classification level description-for a ------General Emergency. BNP Basis Reference(s):

1. OERP BNP Radiological Emergency Response Plan section 3.0 Emergency Response Organization

2. NEI 99-01 HG7--J Deleted: Director I I OPEP-02.2.1 I Rev. 6 1 Page 181 of 310 ATTACHMENT 1 Page 156 of 219 EAL Bases Category S -System Malfunction EAL Group: Hot Conditions (RCS temperature

> 212'F); EALs in this category are applicable only in one or more hot operating modes.Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety.The events of this category pertain to the following subcategories:

1. Loss of Emergency AC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

This category includes loss of onsite and offsite sources for 4160 V emergency buses.2. Loss of Vital DC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of vital plant 125 VDC power sources.3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.

4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (2% -5% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category.

However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits.These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.IOPEP-02.2.1 I Rev. 6 1 Page 182 of 310 ATTACHMENT 1 Page 157 of 219 EAL Bases 5. RCS Leakage The reactor pressure vessel provides a volume for the coolant that covers the reactor core.The reactor pressure vessel and associated pressure piping (reactor coolant system)together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and Primary Containment integrity.

6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS)to initiate and complete reactor scrams. In the plant licensing basis, postulated failures of the RPS to complete a reactor scram comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any trip failure event that does not achieve reactor shutdown.

If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and Primary Containment integrity.

7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

8. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant safety system performance or significant visible damage warrant emergency classification under this subcategory.

IOPEP-02.2.1 Rev. 6 Page 183 of 310 ATTACHMENT 1 Page 158 of 219 EAL Bases Category: S -System Malfunction Subcategory: 1 -Loss of Emergency AC Power Initiating Condition: Loss of all offsite AC power capability to emergency buses for 15 minutes or longer EAL: SU1.1 Unusual Event Loss of all offsite AC power capability to Emergency 4 KV Buses El (E3) and E2(E4) for a 15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None Basis: BNP Basis: The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. 4160 VAC power is divided into Balance of Plant (BOP Bus), and Emergency (E-Bus) distribution. The BOP Buses consist of the Conventional Buses (Common A/B Buses;Buses 1 B/2B, 1 C/2C, 1 D/2D) and Caswell Beach Buses A/B. The Emergency switchgear are Buses El/E3 (Division I) and E2/E4 (Division II).The E-Buses are normally powered through the respective BOP Buses (1D to El, 1 C to E2, 2D to E3, 2C to E4) via a master/slave breaker arrangement. Each E-Bus has a dedicated Diesel Generator to supply an emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of offsite power. The DGs will automatically start and tie onto the E-Buses if the normal power source or offsite power is lost. The DGs can provide power to the E-Buses only. In the event the diesel generator is unavailable for an E-Bus, crosstie capability exists for each E-Bus from the same division of the opposite unit (El to E3, E2 to E4). Although the E-Buses within the unit also have crosstie capability, this alignment is not permitted by plant procedures, with the exception of El to E2 during specific Alternate Safe Shutdown (ASSD) conditions. OPEP-02.2.1 Rev. 6 Page 184 of 310 ATTACHMENT 1 Page 159 of 219 EAL Bases During periods of unit shutdown, when the Startup Auxiliary Transformer (SAT) would be the only normal source of offsite power, the Unit Auxiliary Transformer (UAT) can be made available by establishing a UAT backfeed. Backfeed from the UAT will require the use of keys for the control selector switches and opening of the respective generator's manual no-load disconnect. (Ref. 1, 2)The 15-minute interval was selected as a threshold to exclude transient or momentary power losses.NEI 99-01 Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency buses. This condition represents a potential reduction in the level of safety of the plant.For emergency classification purposes, "capability" means that an offsite AC power source(s)is available to the emergency buses, whether or not the buses are powered from it.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.Escalation of the emergency classification level would be via IC SAI.BNP Basis Reference(s):

1. BNP Updated FSAR Chapter 8 2. 1(2)OP-50 Plant Electric System Operating Procedure 3. NEI 99-01 SU1 I OPEP-02.2.1 I Rev. 6 1 Page 185 of 310 ATTACHMENT 1 Page 160 of 219 EAL Bases Category: Subcategory:

Initiating Condition: S -System Malfunction 1 -Loss of Emergency AC Power Loss of all but one AC power source to emergency buses for 15 minutes or longer EAL: SA1.1 Alert AC power capability to Emergency 4 KV Buses El (E3) and E2(E4) reduced to a single power source for -15 min. (Note 1)AND Any additional single power source failure will result in loss of all unit-specific AC power to SAFETY SYSTEMS Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): 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 (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. OPEP-02.2.1 Rev. 6 Page 186 of 310 ATTACHMENT 1 Page 161 of 219 EAL Bases Basis: BNP Basis: The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. 4160 VAC power is divided into Balance Of Plant (BOP Bus), and Emergency (E-Bus) distribution. The BOP Buses consist of the Conventional Buses (Common A/B Buses;Buses 1 B/2B, 1 C/2C, I D/2D) and Caswell Beach Buses A/B. The Emergency switchgear are Buses E1/E3 (Division I) and E2/E4 (Division II).The E-Buses are normally powered through the respective BOP Buses (1 D to El, 1 C to E2, 2D to E3, 2C to E4) via a master/slave breaker arrangement. Each E-Bus has a dedicated Diesel Generator to supply an emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of offsite power. The DGs will automatically start and tie onto the E-Buses if the normal power source or offsite power is lost. The DGs can provide power to the E-Buses only. In the event the diesel generator is unavailable for an E-Bus, crosstie capability exists for each E-Bus from the same division of the opposite unit (El to E3, E2 to E4). Although the E-Buses within the unit also have crosstie capability, this alignment is not permitted by plant procedures, with the exception of El to E2 during specific Alternate Safe Shutdown (ASSD) conditions. During periods of unit shutdown, when the Startup Auxiliary Transformer (SAT) would be the only normal source of offsite power, the Unit Auxiliary Transformer (UAT) can be made available by establishing a UAT backfeed. Backfeed from the UAT will require the use of keys for the control selector switches and opening of the respective generators manual no-load disconnect (Ref. 1,2).The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. If the capability of a second source of emergency bus power is not restored within 15 minutes, an Alert is declared under this EAL.Because 2 RHR pumps on each unit are powered from the unaffected unit, the words "unit-specific" have been added to clarify that the cross-connected RHR pump power cannot be credited as an AC power source relative to this EAL.NEI 99-01 Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU1.OPEP-02.2.1 Rev. 6 Page 187 of 310 ATTACHMENT 1 Page 162 of 219 EAL Bases I An "AC power source" is a source recognized in AOP anrd_qcapable of supplyin-g reguired ---power to an emergency bus. Some examples of this condition are presented below.* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being 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 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 SS1.BNP Basis Reference(s):

1. BNP Updated FSAR Chapter 8 2. 1(2)OP-50 Plant Electric System Operating Procedure 3. NEI 99-01 SA1-_ Deleted: and EOPs IOPEP-02.2.1 I Rev. 1 Page 188 of 310 ATTACHMENT 1 Page 163 of 219 EAL Bases Category:

S -System Malfunction Subcategory: 1 -Loss of Emergency AC Power Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer EAL: SS1.1 Site Area Emergency Loss of all offsite and all onsite AC power capability to Emergency 4 KV Buses El (E3)and E2(E4) for Z 15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: This EAL is indicated by the loss of all offsite and onsite AC power capability to Emergency 4 KV Buses El (E3) and E2(E4) for greater than or equal to 15 minutes.The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. 4160 VAC power is divided into Balance Of Plant (BOP Bus), and Emergency (E-Bus) distribution. The BOP Buses consist of the Conventional Buses (Common A/B Buses;Buses 1 B/2B, 1 C/2C, 1 D/2D) and Caswell Beach Buses A/B. The Emergency switchgear are Buses E1/E3 (Division I) and E2/E4 (Division II).The E-Buses are normally powered through the respective BOP Buses (1 D to El, 1 C to E2, 2D to E3, 2C to E4) via a master/slave breaker arrangement. Each E-Bus has a dedicated Diesel Generator to supply an emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of offsite power. The DGs will automatically start and tie onto the E-Buses if the normal power source or offsite power is lost. The DGs can provide power to the E-Buses only. In the event the diesel generator is unavailable for an E-Bus, crosstie capability exists for each E-Bus from the same division of the opposite unit (El to E3, E2 to E4). Although the E-Buses within the unit also have crosstie capability, this alignment is not permitted by plant procedures, with the exception of El to E2 during specific Alternate Safe Shutdown (ASSD) conditions. OPEP-02.2.1 Rev. 6 Page 189 of 310 ATTACHMENT 1 Page 164 of 219 EAL Bases During periods of unit shutdown, when the Startup Auxiliary Transformer (SAT) would be the only normal source of offsite power, the Unit Auxiliary Transformer (UAT) can be made available by establishing a UAT backfeed. Backfeed from the UAT will require the use of keys for the control selector switches and opening of the respective generator's manual no-load disconnect (Ref. 1, 2)The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. The interval begins when both offsite and onsite AC power are lost.NEI 99-01 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 ICsBRG1, FG1 or SG1.BNP Basis Reference(s): , J Deleted: AG1]1.2.3.4.BNP Updated FSAR Chapter 8 1(2)OP-50 Plant Electric System Operating Procedure OAOP-36.2 Station Blackout NEI 99-01 SS1 I OPEP-02.2.1 Rev. 6 1 Page 190 of 310 1 ATTACHMENT 1 Page 165 of 219 EAL Bases S -System Malfunction 1 -Loss of Emergency AC Power Category: Subcategory: Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses OR loss of all emergency AC and vital DC power sources for 15 minutes or longer EAL: SG1.1 General Emergency Loss of all offsite and all onsite AC power capability to Emergency 4 KV Buses El (E3)and E2(E4)AND EITHER: " Restoration of at least one emergency bus in < 4 hours is not likely (Note 1)" RPV water level cannot be restored and maintained > MSCRWL (LL-4)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: This EAL is indicated by the extended loss of all offsite and onsite AC power capability to 4 KV emergency buses El (E3) and E2(E4) either for greater then the BNP Station Blackout (SBO)coping analysis time (4 hrs.) (ref. 1) or that has resulted in indications of an actual loss of adequate core cooling.Indication of continuing core cooling degradation is manifested by the inability to restore and maintain RPV water level above the Minimum Steam Cooling Reactor Water Level (LL-4) (ref.4, 5).I OPEP-02.2.1 Rev. 6 1 Page 191 of 310 ATTACHMENT 1 Page 166 of 219 EAL Bases The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. 4160 VAC power is divided into Balance of Plant (BOP Bus), and Emergency (E-Bus) distribution. The BOP Buses consist of the Conventional Buses (Common A/B Buses;Buses 1 B/2B, 1 C/2C, 1 D/2D) and Caswell Beach Buses A/B. The Emergency switchgear are Buses E1/E3 (Division I) and E2/E4 (Division II).The E-Buses are normally powered through the respective BOP Buses (1 D to El, 1 C to E2, 2D to E3, 2C to E4) via a master/slave breaker arrangement. Each E-Bus has a dedicated Diesel Generator to supply an emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of offsite power. The DGs will automatically start and tie onto the E-Buses if the normal power source or offsite power is lost. The DGs can provide power to the E-Buses only. In the event the diesel generator is unavailable for an E-Bus, crosstie capability exists for each E-Bus from the same division of the opposite unit (El to E3, E2 to E4). Although the E-Buses within the unit also have crosstie capability, this alignment is not permitted by plant procedures, with the exception of El to E2 during specific Alternate Safe Shutdown (ASSD) conditions. During periods of unit shutdown, when the Startup Auxiliary Transformer (SAT) would be the only normal source of offsite power, the Unit Auxiliary Transformer (UAT) can be made available by establishing a UAT backfeed. Backfeed from the UAT will require the use of keys for the control selector switches and opening of the respective generator's manual no-load disconnect (Ref. 2, 3).Four hours is the station blackout coping time (ref 1).Indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Site Emergency Coordinator judgment as it relates to imminent Loss or Potential Loss of fission product barriers and degraded ability to monitor fission product barriers. Indication of continuing core cooling degradation is manifested by an RPV level that cannot be restored and maintained > MSCRWL (LL-4) (ref. 4, 5). Core submergence is the most desirable means of core cooling, however when RPV level is below TAF, the uncovered portion of the core can be cooled by less reliable means (i.e., steam cooling or spray cooling).NEI-9901 Basis: This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. OPEP-02.2.1 Rev. 6 Page 192 of 310 ATTACHMENT 1 Page 167 of 219 EAL Bases The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions.Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of 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 bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.BNP Basis Reference(s): 1.2.3.4.5.OAOP-36.2 STATION BLACKOUT, Section 4.0 BNP Updated FSAR Chapter 8 1(2)OP-50 Plant Electric System Operating Procedure 1(2)EOP-01 Reactor Vessel Control Procedure OEOP-01-NL, EOP SAMG NUMERICAL LIMITS AND VALUES I OPEP-02.2.1 I Rev. 6 1 Page 193 of 310 ATTACHMENT 1 Page 168 of 219 EAL Bases S -System Malfunction 1 -Loss of Emergency AC Power Category: Subcategory: Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses OR loss of all emergency AC and vital DC power sources for 15 minutes or longer EAL: SG11.2 General Emergency Loss of all offsite and all onsite AC power capability to Emergency 4 KV Buses El (E3)and E2(E4) for > 15 min.AND Loss of all 125 VDC power based on battery bus voltage indications < 105 VDC on all vital DC buses 1(2)A-1, A-2, B-1 and B-2 for> 15 min.(Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: This EAL is indicated by the loss of all offsite and onsite emergency AC power capability to 4 KV emergency buses El (E3) and E2(E4) for greater than 15 minutes in combination with degraded vital DC power voltage. This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi.The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. 4160 VAC power is divided into Balance of Plant (BOP Bus), and Emergency (E-Bus) distribution. The BOP Buses consist of the Conventional Buses (Common A/B Buses;Buses 1 B/2B, 1 C/2C, 1 D/2D) and Caswell Beach Buses A/B. The Emergency switchgear are Buses E1/E3 (Division I) and E2/E4 (Division II).OPEP-02.2.1 Rev. 6 Page 194 of 310 ATTACHMENT 1 Page 169 of 219 EAL Bases The E-Buses are normally powered through the respective BOP Buses (1 D to El, 1 C to E2, 2D to E3, 2C to E4) via a master/slave breaker arrangement. Each E-Bus has a dedicated Diesel Generator to supply an emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of offsite power. The DGs will automatically start and tie onto the E-Buses if the normal power source or offsite power is lost. The DGs can provide power to the E-Buses only. In the event the diesel generator is unavailable for an E-Bus, crosstie capability exists for each E-Bus from the same division of the opposite unit (El to E3, E2 to E4). Although the E-Buses within the unit also have crosstie capability, this alignment is not permitted by plant procedures, with the exception of El to E2 during specific Alternate Safe Shutdown (ASSD) conditions. During periods of unit shutdown, when the Startup Auxiliary Transformer (SAT) would be the only normal source of offsite power, the Unit Auxiliary Transformer (UAT) can be made available by establishing a UAT backfeed. Backfeed from the UAT will require the use of keys for the control selector switches and opening of the respective generator's manual no-load disconnect (Ref. 2, 3).There are two independent vital 125 VDC divisions per unit, designated Division I and Division II (Batteries 1 (2)A-1 and 1(2)A-2 for Division I and Batteries 1 (2)B-1 and 1(2)B-2 for Division II). Each division consists of a 250 VDC battery center tapped to form two 125 VDC batteries. Each 125 VDC battery has an associated full capacity battery charger. The chargers are supplied from the same AC load groups for which the associated DC subsystem supplies the control power. During normal operation, the DC loads are powered from the battery chargers with the batteries floating on the system. In case of loss of normal power to the battery charger, the DC loads are automatically powered from the station batteries. 105 VDC is the minimum design voltage limit (ref. 4, 5).NEI-9901 Basis: This IC addresses a concurrent and prolonged loss of both emergency AC and Vital DC power. A loss of all emergencyAC 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 emerqency AC and vital DC power will lead to multiple challenges to fission product barriers.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.IOPEP-02.2.1 I Rev. 6 Page 195 of 310o ATTACHMENT 1 Page 170 of 219 EAL Bases BNP Basis Reference(s): 1.2.3.4.5.OAOP-36.2 STATION BLACKOUT, Section 4.0 BNP Updated FSAR Chapter 8 1(2)OP-50 Plant Electric System Operating Procedure BNP Technical Specification Bases B.3.8.4 OAOP-39.0 LOSS OF DC POWER I0PEP-02.2.1 I Rev. 6 Page 196 of 310 ATTACHMENT 1 Page 171 of 219 EAL Bases Category: S -System Malfunction Subcategory: 2 -Loss of Vital DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer EAL: SS2.1 Site Area Emergency Loss of all 125 VDC power based on battery bus voltage indications < 105 VDC on all vital DC buses 1 (2)A-1, A-2, B-1 and B-2 for 2 15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: There are two independent vital 125 VDC divisions per unit, designated Division I and Division II (Batteries 1(2)A-1 and 1(2)A-2 for Division I and Batteries 1 (2)B-1 and 1(2)B-2 for Division II). Each division consists of a 250 VDC battery center tapped to form two 125 VDC batteries. Each 125 VDC battery has an associated full capacity battery charger. The chargers are supplied from the same AC load groups for which the associated DC subsystem supplies the control power. During normal operation, the DC loads are powered from the battery chargers with the batteries floating on the system. In case of loss of normal power to the battery charger, the DC loads are automatically powered from the station batteries. Note that the Control Room DC voltage indicator only reads battery charger output voltage and not battery voltage unless the charger output breaker is closed. However ERFIS does provide DC battery voltage, otherwise battery voltage must be read locally.105 VDC is the minimum design voltage limit (ref. 1).NEI 99-01 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.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.0PEP-02.2.1 Rev. 6 Page 197 of 310 Escalation of the emergency classification level would be via ICs FG1, Fi or,$G1._ Deleted: AG1 I.... ......... --, -_ Deleted: SG8 I IOPEP-02.2.1 I Rev. 6 Page 198 of 310 ATTACHMENT 1 Page 173 of 219 EAL Bases BNP Basis Reference(s):

1. BNP Technical Specification Bases B.3.8.4 2. OAOP-39.0 LOSS OF DC POWER 3. NEI 99-01 SS8 OPEP-02.2.1 Rev. 6 Page 199 of 310 ATTACHMENT 1 Page 174 of 219 EAL Bases S -System Malfunction 3 -Loss of Control Room Indications Category: Subcategory:

Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL: SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for > 15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table S-1 Safety System Parameters" Reactor power* RPV water level" RPV pressure* Primary containment pressure* Torus water level* Torus temperature Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): 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.BNP Basis: SAFETY SYSTEM parameters listed in Table S-1 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.The ERFIS, which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1, 2).As used in this EAL "within the Control Room" means any available indicator available within the Control Room boundary, including back panels.OPEP-02.2.1 IRev. 6 Page 200 of 310 ATTACHMENT 1 Page 175 of 219 EAL Bases NEI 99-01 Basis: This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital or-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-1 022) 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. _ -j Deleted: and I This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV leveljradRCS 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 forRPV -_water levelcannot 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.Deleted: core cooling [PWR] /Deleted: [BWRF_ --[ Deleted: reactor vessel level [PWR] /Deleted: [BWR], 4 Deleted: SA2 Escalation of the emergency classification level would be via IC,.A3 -.---------------- -BNP Basis Reference(s):

1. Updated FSAR Update Section 7.7.1.9 2. 001-01.08 Control of Equipment and System Status 3. NEI 99-01 SU2 IOPEP-02.2.1 ev. 1 Page 201 of 310 ATTACHMENT 1 Page 176 of 219 EAL Bases Category:

S -System Malfunction Subcategory: 3 -Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL: SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for >_ 15 min. (Note 1)AND Any significant transient is in progress, Table S-2 Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table S-1 Safety System Parameters

Reactor power" RPV water level* RPV pressure* Primary containment pressure* Torus water level" Torus temperature Table S-2 Significant Transients
Reactor scram" Runback > 25% rated thermal power* Electrical load rejection

> 25%electrical load" ECCS injection" Thermal power oscillations > 10%(peak to peak)I OPEP-02.2.1 Rev. 6 1 Page 202 of 310 ATTACHMENT 1 Page 177 of 219 EAL Bases Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): 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.BNP Basis: SAFETY SYSTEM parameters listed in Table S-1 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.The ERFIS, which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1, 2).Significant transients are listed in Table S-2 and include response to automatic or manually initiated functions such as scrams, runbacks (Recirculation) involving greater than 25%thermal power change, electrical load rejections of greater than 25% full electrical load, ECCS injections, or thermal power oscillations of 10% (peak to peak) or greater.As used in this EAL "within the Control Room" means any available indicator available within the Control Room boundary, including back panels.NEI 99-01 Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor I power level cannot be determined from any analog, digital or recorder source within the ---- Deleted: and 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-1 022) 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. OPEP-02.2.1 IRev. 6 1Page 203 of 310 I ATTACHMENT 1 Page 178 of 219 EAL Bases This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV leveland 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 forJRPV water level pqannot 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 FS1 or ICBRS1 BNP Basis Reference(s):

1. Updated FSAR Update Section 7.7.1.9 2. 001-01.08 Control of Equipment and System Status 3. NEI 99-01 SA2 Deleted: core cooling [PWR] /Deleted: [BWql--Deleted: reactor vessel level [PWRq /JDeleted:

13BWR , j Deleted: AS1 IOPEP-02.2.1 I Rev. 6 Page 204 of 310 ATTACHMENT 1 Page 179 of 219 EAL Bases S -System Malfunction Category: Subcategory: 4 -RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL: SU4.1 Unusual Event Steam Jet Air Ejector Radiation Monitor 1(2)D12-RM-K601 A /B Hi-Hi alarm (Process Off-Gas Rad Hi-Hi alarm 1(2)UA-03 4-2) > 15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, orVill likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: The Steam Jet Air Ejector radiation monitor setpoint provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10CFR100 in the event of an inadvertent release via the condenser air ejector (ref. 2, 3).At the Hi-Hi alarm setpoint, the process Off-Gas timer is started. After the process Off-Gas timer has timed out (15 minutes), the Off-Gas system will isolate (ref. 1).NEI 99-01 Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.Escalation of the emergency classification level would be via ICs FA1 or the Recognition C a te g o ry .lC s -. ----------------------------------------------------Deleted: A OPEP-02.2.1 Rev. 6 Page 205 of 310 ATTACHMENT 1 Page 180 of 219 EAL Bases BNP Basis Reference(s):

1. ARP 1(2)APP-UA-03 4-2 Process Off-Gas Rad Hi-Hi 2. BNP Offsite Dose Calculation Manual section 3.1.3 3. BNP Technical Specifications section 3.7.5 4. NEI 99-01 SU3 I OPEP-02.2.1 Rev. 6 1 Page 206 of 310 ATTACHMENT 1 Page 181 of 219 EAL Bases S -System Malfunction Category: Subcategory:

4 -RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL: SU4.2 Unusual Event Coolant activity > 0.2 pCi/gm 1-131 dose equivalent for > 48 hours OR Coolant activity > 4.0 pCi/gm 1-131 dose equivalent instantaneous Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: The specific iodine activity is limited to -0.2 VCi/gm Dose Equivalent 1-131. This limit ensures the source term assumed in the safety analysis for the Main Steam Line Break (MSLB) is not exceeded, so any release of radioactivity to the environment during an MSLB is less than a small fraction of the 10 CFR 50.67 limits (ref. 1).The upper limit of 4.0 pCi/gm Dose Equivalent 1-131 ensures that the thyroid dose from an MSLB will not exceed the dose guidelines of 10 CFR 50.67 or Control Room operator dose limits specified in GDC 19 of 10 CFR 50, Appendix A (ref. 1).NEI 99-01 Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category.R ICss. ....................................................... --4 Deleted: A I OPEP-02.2.1 I Rev. 6 1 Page2of 310 ATTACHMENT 1 Page 182 of 219 EAL Bases BNP Basis Reference(s):

1. BNP Technical Specifications section 3.4.6 2. NEI 99-01 SU3 IOPEP-02.2.1 I Rev. 6 Page 208 of 310 ATTACHMENT 1 Page 183 of 219 EAL Bases Category:

S -System Malfunction Subcategory: 5 -RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL: SU5.1 Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for > 15 min.OR RCS identified leakage > 25 gpm for > 15 min.OR Leakage from the RCS to a location outside Primary Containment > 25 gpm for > 15 min.(Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: Leakage is monitored by utilizing the following techniques: " Sensing excess flow in piping systems" Sensing pressure and temperature changes in the primary containment" Monitoring for high flow and temperature through selected drains," Sampling airborne particulate and gaseous radioactivity.

Drywell floor and equipment drain sump leak rate system I OPEP-02.2,1 Rev. 6 1 Page 209 of 310 ATTACHMENT 1 Page 184 of 219 EAL Bases Identified leakage is leakage into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or leakage into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary leakage. (ref.1,2)Unidentified leakage is all leakage into the drywell that is not identified leakage. (ref. 1, 2)Pressure boundary leakage is leakage through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall. (ref. 1, 2)The drywell floor drain sump flow monitoring system monitors the leakage collected in the floor drain sump. This unidentified leakage consists of leakage from control rod drives, valve flanges, floor drains, the Reactor Building Closed Cooling Water System, and drywell cooler drains, and any leakage not collected in the drywell equipment drain sump. The drywell floor drain sump is provided with two sump pumps. A flow transmitter in the common discharge line of the drywell floor drain sump pumps inputs to a flow integrator.

In addition to the required instrumentation, the starting frequency and run duration of a sump pump motor are monitored by timer circuitry to provide a signal (alarm) in the Control Room indicating that leakage has reached a specified limit. (ref. 2)RCS leakage outside of the Primary Containment that is not considered identified or unidentified leakage per Technical Specifications includes leakage via interfacing systems such as RCS to the Reactor Building Closed Cooling Water (RBCCW system), or systems that directly see RCS pressure outside primary containment such as Reactor Water Cleanup (RWCU), reactor water sampling system and Residual Heat Removal (RHR) system (when in the shutdown cooling mode) (ref. 3)Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation, EAL FA1.1. The note has been added to remind the EAL-user to review Table F-1 for possible escalation to higher emergency classifications. NEI 99-01 Basis: This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.IOPEP-02.2.1 I Rev. 6 Page 210 of 310 ATTACHMENT 1 Page 185 of 219 EAL Bases.The first and second EAL conditions are focused on a loss of mass from the RCS due to- --Deleted: EAL#1 and EAL#2"unidentified leakage", "pressure boundary leakage' or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). The third condition addresses -san --- --- Deleted: EAL #3 RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into thePrimary Containment or a location outside-of Primary Deleted: EALs Containment. ,. "Deleted: containment 1 Deleted: a secondary-side system (e.g., The leak rate values for each condition were selected because they are usually_observable steam generator tube leakage in a PWR)with normal Control Room indications. Lesser values typically require time-consuming Deleted: containment calculations to determine (e.g., a mass balance calculation). _The first condition uses a lower Deleted: EAL value that reflects the greater significance of unidentified or pressure boundary leakage. Deleted: EAL #1 Deleted: The release of mass from the RCS ,_ stuck-open Safety Relief Valve (SRV)Aor_SRV leakage is not considered either identified or due -designed/expcted operation of a unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL. -relief valve does not warrant an emergency classification. For PWRs, an emergency-classification would be required if a mass loss The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate \ is caused by a relief valve that is not functioning the leakage, if possible. as designed/expected (e.g., a relief valve sticks open an e ne ow cann- e so e .Escalation of the emergency classification level would be via ICs of Recognition Category_.or _ _ le For BWRs, a F. A Deleted: A BNP Basis Reference(s):

1. BNP Technical Specifications Definitions section 1.1 2. BNP Technical Specifications Bases 3.4.5 3. BNP UFSAR section 5.1 Reactor Coolant System and Connected Systems 4. NEI 99-01 SU4 IOPEP-02.2.1 I Rev. 6 Page 211 of 310o ATTACHMENT 1 Page 186 of 219 EAL Bases S -System Malfunction Category: Subcategory:

6- RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL: SU6.1 Unusual Event An automatic scram did not reduce reactor power to < 2% (APRM downscale) after any RPS setpoint is exceeded AND A subsequent automatic scram or manual scram action taken at the reactor control console (Manual PBs, Mode Switch, ARI) is successful in shutting down the reactor as indicated by reactor power < 2% (APRM downscale) (Note 8)Note 8: A manual scram 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. Mode Applicability: 1 -Power Operations, 2 -Startup Definition(s): None BNP Basis: The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protection System (RPS) scram function. A reactor scram is automatically initiated by the Reactor Protection System (RPS) when certain continuously monitored parameters exceed predetermined setpoints (ref. 1).Following a successful reactor scram, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative period. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-scram response from an automatic reactor scram signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful scram has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the APRM downscale setpoint of 2%.OPEP-02.2.1 IRev. 6 Page 212 of 310 ATTACHMENT 1 Page 187 of 219 EAL Bases For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., manual scram pushbuttons, mode switch, or ARI actuation). Reactor shutdown achieved by use of the 0EOP-01-LEP-02 actions does not constitute a successful manual scram (ref. 4).Following any automatic RPS scram signal, OEOP-01 (ref. 2) and OEOP-01 -LPC (ref. 3)prescribe insertion of redundant manual scram signals to back up the automatic RPS scram function and ensure reactor shutdown is achieved. Even if the first subsequent manual scram signal inserts all control rods to the full-in position immediately after the initial failure of the automatic scram, the lowest level of classification that must be declared is an Unusual Event.Taking the mode switch to shutdown is a manual scram action. When the Mode Switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered.If reactor power remains above the lowered setpoint, an automatic scram is initiated. For the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power below 2% is not considered a successful automatic scram. If automatic actuation of ARI has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. However, a successful automatic or manual initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions.In the event that the operator identifies a reactor scram is imminent and initiates a successful manual reactor scram before the automatic scram setpoint is reached, no declaration is required. The successful manual scram of the reactor before it reaches its automatic scram setpoint or reactor scram signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. If manual reactor scram actions fail to reduce reactor power below 2% (ref. 2, 3), the event escalates to the Alert under EAL SA6.1.If by procedure, operator actions include the initiation of an immediate manual scram following receipt of an automatic scram signal and there are no clear indications that the automatic scram failed (such as a time delay following indications that a scram setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic scram or manual actions. If a subsequent review of the scram actuation indications reveals that the automatic scram did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 50.72 should be considered for the-transient event.IOPEP-02.2.1 I Rev. 6 1 Page 213 of 310 ATTACHMENT 1 Page 188 of 219 EAL Bases NEI 99-01 Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor ,cram that results in a reactor shutdown, and either a subsequent operator manual action ---Deleted: (trip [PWR] /taken-at th1e reactor control consoles or an automatic scrarn, is -successful -in shutting down the Deleted: [BW)reactor. This event is a precursor to a more significant condition and thus represents a e. (trip [PWR /potential degradation of the level of safety of the plant. '(Deleted: [BWR])Following the failure on an automatic reactor_scranr operators willoroptlv initiate manual -" l p................................... a--a -Deleted: [BWR])actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor -: ,scranr.. If-these manual actions are successful in shutting d own the reactor, core heat ---Deleted: (trip [PWR] /generation will quickly fall to a level within the capabilities of the plant's decay heat removal Deleted: [BWRJ)systems.If an initial manual reactorscrarm, is unsuccessful, operators will promptly take manual action .Deleted: (trip [PWF /at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a Deleted: [BWR)manual reactor scrakn'using a different switch). Dependingupon several factors, the initial or --Deleted: (trip [PWR]/subsequent effort to manuallyscram ,the reactor, or a concurrent plant condition, m lead to Deleted: [BWR])the generation of an automatic reactorscra_,signal. If a subsequent manual or automatic Deleted: (trip [PWq /scram_,is successful in shutting down the reactor, core heat generation will quickly fall to a \ --eleted:[w") level within the capabilities of the plant's decay heat removal systems. ','"Deleted: (trip [PWR]I A manual action at the reactor control consoles is any operator action, or set of actions, which , Deleted: [BWFR)causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor ,scaran. ThPis action does not include manually drivvingin control rods or impjementation 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". \ Deleted: (trip EPWRJ /IDeleted: [BWR()Deleted: (trip [PWR]'fDeleted: [BWFJ)Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.,_ _, -Deleted: [BWRJ I f The plant response to the failure of an automatic or manual reactorciran_,wil vary based --Deleted: (trip [PWR]/upon several factors including the reactor power level prior to the event, availability of the -Deleted: [BWR])condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via ICSA6._ Depending upon thje plant response, escalation is also ----- Deleted: SA5 possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 orFAl, an -- Deleted: SA5 Unusual Event declaration is appropriate for this event., I Deleted: A reactor shutdown is determined in'/accordance with applicable Emergency Operating Procedure criteria.ATTACHMENT 1 Page 188 of 219 EAL Bases IOPEP-02.2.1 I Rev. 6 1 Page 214 of 310 1 Should a reactorsqcranjsi gnal 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 scra@Tand the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated. e If the signal does not cause a plant transient and the scrar,_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. i Deleted: (trip [PWR] /Deleted: [BWR])Deleted: (trip [PWR] /Deleted: [BWR]).[Deleted: (trip [PWR] /--I Deleed:[BWR]) BNP Basis Reference(s):

1. BNP Technical Specifications section 3.3.1.1 RPS Instrumentation

2. 1(2) EOP-01, Reactor Scram Procedure 3. 1(2) EOP-01 -LPC, Level/Power Control 4. OEOP-01-LEP-02 Alternate Control Rod Insertion 5. NEI 99-01 SU5 OPEP-02.2.1 Rev.6 1 Page 25of 310 ATTACHMENT 1 Page 190 of 219 EAL Bases S -System Malfunction 6- RPS Failure Category: Subcategory:

Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL: SU6.2 Unusual Event A manual scram did not reduce reactor power to < 2% (APRM downscale) after any manual scram action was initiated AND A subsequent automatic scram or manual scram action taken at the reactor control console (Manual PBs, Mode Switch, ARI) is successful in shutting down the reactor as indicated by reactor power < 2% (APRM downscale) (Note 8)Note 8: A manual scram 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. Mode Applicability: 1 -Power Operations, 2 -Startup Definition(s): None BNP Basis: This EAL addresses a failure of a manually initiated scram in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual scram is successful in shutting down the reactor (reactor power < 2%). (ref. 1).Following a successful reactor scram, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative period. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-scram response from a manual reactor scram signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful scram has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the APRM downscale setpoint of 2%.ATTACHMENT 1 OPEP-02.2.1 Rev. 6 Page 216 of 310 Page 191 of 219 EAL Bases For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., manual scram pushbuttons, mode switch, or ARI actuation). Reactor shutdown achieved by use of the OEOP-01-LEP-02 actions does not constitute a successful manual scram (ref. 2, 3).Taking the mode switch to shutdown is a manual scram action. When the Mode Switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered.If reactor power remains above the lowered setpoint, an automatic scram is initiated. Successful automatic or manual initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions.If both subsequent automatic and subsequent manual reactor scram actions in the Control Room fail to reduce reactor power below the power associated with the safety system design (< 2%) following a failure of an initial manual scram, the event escalates to an Alert under EAL SA6.1 NEI 99-01 Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor ,scramthat results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scra., 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.--(Deleted: (trip [PWRM /Deleted: [BWFR)", ( Deleted: (trip [PWR]/{Deleted: [BWR]) }p .......... .p ....eezg .ra .[ ..W ..Following the failure on an automatic reactor _scra_n,_ operators wilpromptly iritiate manual --(actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor .... -[ -,scrarn(./f tIhese manual-actions are successful in shutting down the reactor, core heat -, Deleted: (tnp [PW/generation will quickly fall to a level within the capabilities of the plant's decay heat removal .< Deleted: [BWL])systems.I ~-.-~--.-If an initial manual reactorscrarm, is unsuccessful, operators will promptly take manual action --uemee: at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a Deleted: [BWR)manual reactorsqra _ usin qa different switch). Depending upon several factors, the initial or --- Deleted: (trip [OWR]/subsequent effort to manuallyjcram, the reactor, or a concurrent plant condition, may lead to 1 Deleted: [BWRI)the generation of an automatic reactorAscram, signal. If a subsequent manual or automatic --- Deleted: (trp [PWR ,scram, is successful in shutting down the reactor, core heat_generation will gic-kly-fall to a Deleted: [BW])level within the capabilities of the plant's decay heat removal systems. ,al jDeleted:[ (WR]Deleted: (tdp [PWR] I\\Deleted: [BWRJ)j 3 3 Deleted: (trip [PWR] /Deleted: [BWR])IOPEP-02.2.1 I Rev. 6 Page217of 310 ATTACHMENT 1 Page 192 of 219 EAL Bases A manual action at the reactor control consoles 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". -Deleted: (trip [PWRI]Deleted: [BWR])--i 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 reactorcramn,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 are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC.SA6. Depending upon the plant response, escalation is also possible via IC FAI. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.-Deleted: [BWR]Deleted: (trip [PWR] /Deleted: [BWR])--Deleted: SA5 I -[ Deleted: SA5 I j}~I Deleted: A reactor atii Should -a reactor scrasig~n~al be generated as a result of plant work (_e.g., RPS-setpoint -- -accordance with appticr testing), the following classification guidance should be applied. Operating Procederting\J Deleted: (trip [PWR] /* If the signal causes a plant transient that should have included an automatic reactor t Deleted: [BWR])sqrarmnand the RPS fails to-automatically-shutdown the reactor, then this IC and the --- -Deleted: (trip[PWR/ EALs are applicable, and should be evaluated. Deleted: -pWR]/Deleted: [BWR])utdown is determined in able Emergency iteria.¶9 If the signal does not cause a plant transient and thepscrarnfailure is determined -ueleteea: (tnp [PW through other means (e.g., assessment of test results), th-en-this-1IC-and-t-he-E-AL-sare ..Deleted: [BWR])not applicable and no classification is warranted. t I/BNP Basis Reference(s):

1. BNP Technical Specifications section 3.3.1.1 RPS Instrumentation

2. 1(2) EOP-01, Reactor Scram Procedure 3. OEOP-01 -LEP-02 Alternate Control Rod Insertion 4. NEI 99-01 SU5 OPEP-02.2.1 Rev. 6 1 Page 218 of 310 1 Category: Subcategory:

Initiating Condition: ATTACHMENT 1 Page 193 of 219 EAL Bases S -System Malfunction 2 -RPS Failure Automatic or manual scram fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL: SA6.1 Alert An automatic or manual scram fails to reduce reactor power to < 2% (APRM downscale) AND Manual scram actions taken at the reactor control console (Manual PBs, Mode Switch, ARI) are not successful in shutting down the reactor as indicated by reactor power > 2%(Note 8)Note 8: A manual scram 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. Mode Applicability: 1 -Power Operations, 2 -Startup Definition(s): None BNP Basis: This EAL addresses any automatic or manual reactor scram signal that fails to shut down the reactor followed by a subsequent manual scram that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed.For the purposes of emergency classification at the Alert level, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., manual scram pushbuttons, mode switch, or ARI actuation). Reactor shutdown achieved by use of the OEOP-01 -LEP-02 actions does not constitute a successful manual scram (ref. 1).I OPEP-02.2.1 Rev. 6 1 Page 219 of 310 ATTACHMENT 1 Page 194 of 219 EAL Bases For the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power below 2% is not considered a successful automatic scram. If automatic actuation of ARI has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. However, a successful automatic initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions.The APRM downscale trip setpoint (2%) is a minimum reading on the power range scale that indicates power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM)indications or other reactor parameters (steam flow, RPV pressure, torus temperature trend)can be used to determine if reactor power is greater than 2 % power (ref. 2, 3).Escalation of this event to a Site Area Emergency would be under EAL SS6.NEI 99-01 Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor sqcr~amthat results in a reactor shutdown, and subsequen~t operator manual actions taken at -. Deleted: (trip [PWR] /the reactor control consoles to shutdown the reactor are also unsuccessful. This condition -Deleted: [BWR])represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.A manual action at the reactor control console 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, -- ----Deleted: (trip [PWRJ /This action does not include manually driving in control rods or implementation of --Deleted: [BWR])boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console,(e.g., -- -- -Dlted: s locally opening breakers). Actions taken at backpanels or other locations within the Control _ --- Deleted: -Room, or any location outside the Control Room, are not considered to be "at the reactor control console." ....---------------------------------------------- -_ --Deleted: s Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action _, --foDeleted: [BWRJ I OPEP-02.2.1 I Rev. 6 I Page 220 of 310 ATTACHMENT 1 Page 195 of 219 EAL Bases f -_,_LJ_ .The plant response to the failure of an automatic or manual reactorcrarwill vary based 1uelete: (tr upon several factors including the reactor power level prior to the event, availability of the -- -- Deleted:-[ condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge toRPV water levej, or RCS heat removal safety functions, the emergency .. -.. Deleted: th classification level will escalate to a-Site Area Emergency via IC SS4 Depending upon plant ... -Deleted:[responses and symptoms, escalation is also possible via IC FSI. Absent the plant conditions Deleted: 5 needed to meet either IC SS.6,or FS1, an Alert declaration is appropriate-for this event ......... -1 Deleted: 5 ip [PWR] /BWRI)e or oolng (PWR I BWRR!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. BNP BasisReference(s)-, I Deleted: A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.¶1 1.2.3.4.OEOP-01-LEP-02, Alternate Control Rod Insertion 1(2) EOP-01, Reactor Scram Procedure 1(2) EOP-01 -LPC, Level/Power Control NEI 99-01 SA5 OPEP-02.2.1 Rev. 6 Page 221 of 310 ATTACHMENT 1 Page 196 of 219 EAL Bases Category: S -System Malfunction Subcategory: 2 -RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to RPV water level or RCS heat removal EAL: SS6.1 Site Area Emergency An automatic or manual scram fails to reduce reactor power to < 2% (APRM downscale) AND All actions to shut down the reactor are not successful as indicated by reactor power> 2%AND EITHER: " RPV level cannot be restored and maintained > LL-4 or cannot be determined" Suppression pool water temperature and RPV pressure cannot be maintained below the HCTL Mode Applicability: 1 -Power Operations, 2 -Startup Definition(s): None BNP Basis: This EAL addresses the following:

Any automatic reactor scram signal followed by a manual scram that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed (EAL SA6.1), and* Indications that either core cooling is extremely challenged or heat removal is extremely challenged.

IOPEP-02.2.1 I Rev. 6 Page 222 of 310 ATTACHMENT 1 Page 197 of 219 EAL Bases Reactor shutdown achieved by use of OEOP-01 -LEP-02 Alternate Control Rod Insertion is also credited as a successful manual scram provided reactor power can be reduced below the APRM downscale trip setpoint before indications of an extreme challenge to either core cooling or heat removal exist.The APRM downscale trip setpoint (2%) is a minimum reading on the power range scale that indicates power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM)indications or other reactor parameters (steam flow, RPV pressure, torus temperature trend)can be used to determine if reactor power is greater than 2% power (ref. 1, 2).The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS barriers.Indication that core cooling is extremely challenged is manifested by inability to restore and maintain RPV water level above LL-4. LL-4 is the Minimum Steam Cooling RPV Water Level (MSCRWL) (ref. 3). The MSCRWL is the lowest RPV level at which the covered portion of the reactor core will generate sufficient steam to prevent any clad temperature in the uncovered part of the core from exceeding 1500 0 F. This water level is utilized in the EOPs to preclude fuel damage when RPV level is below the top of active fuel. RPV level below the MSCRWL for an extended period of time without satisfactory core spray cooling could be a precursor of a core melt sequence. When RPV level cannot be determined, EOPs require entry to EOP-01 -RXFP, Reactor Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained. When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EOP-01 -RXFP specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Alternate Flooding Pressure (ref. 4).The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool water temperature from which Emergency RPV Depressurization will not raise suppression chamber pressure above Primary Containment Pressure Limit A (PCPL-A), while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.The HCTL is a function of RPV pressure and torus water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant. This threshold is met when the final step SP/T-1 3 of section SP/T in EOP-02-PCCP, Primary Containment Control, is reached (ref. 5). This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature. OPEP-02.2.1 I Rev. 6 1 Page 223 of 310 ATTACHMENT 1 Page 198 of 219 EAL Bases NEI 99-01 Basis: J Deleted: ttd, rpiAlm /}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 -- --eeterd: [BW])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. VVJ IJ J J 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.J7scalation of the emergency classification level would be via ICBRG1 or FGI ............. BNP Basis Reference(s): -Deleted: A reactor shutdown is determined in accordance with applicable Emergency I Operating Procedure criteria.¶ Deleted: AG1 1. EOP-01 Reactor Scram Procedure 2. EOP-01 -LPC Level/Power Control 3. 0EOP-01-NL EOP-SAMG Numerical Limits and Values, Attachment 1, pg 37-40, Figures 1-10 and 1-11 4. EOP-01-RXFP, Reactor Flooding 5. EOP-02-PCCP, Primary Containment Control 6. NEI 99-01 SS5 IOPEP-02.2.1 I Rev. 6 1 Page 224 of 310 ATTACHMENT 1 Page 199 of 219 EAL Bases Category: S -System Malfunction Subcategory: 7 -Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: SU7.1 Unusual Event Loss of all Table S-3 onsite communication methods OR Loss of all Table S-3 offsite communication methods OR Loss of all Table S-3 NRC communication methods Table S-3 Communication Methods System Onsite Offsite NRC Public Address System X PBX Telephone System X X X Corporate Telephone X X X Communications System Commercial Telephones X X X Satellite Phones X X Cellular Phones X X NRC Emergency X Telecommunications System OPEP-02.2.1 Rev. 6 Page 225 of 310 ATTACHMENT 1 Page 200 of 219 EAL Bases Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: Onsite/offsite/NRC communications include one or more of the systems listed in Table S-3 (ref. 1).Public Address System The Brunswick Plant public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature. This system is powered from the plant uninterruptible power supply which employs battery reserve as well as diesel generator emergency supply.PBX Telephone System The Brunswick Site PBX telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code. The PBX telephone system also provides for outside communications. The PBX switch located in the TSC/EOF building is also backed up by a battery UPS capable of supplying power for a minimum of 8 hours and is augmented by a Diesel Generator capable of supplying power to the TSC/EOF building for at least 5 days.Corporate Telephone Communications System (Voicenet and/or DEMNET)Interconnected through the site PBX and the emergency telephone system, the Voicenet system provides a means to communicate with other corporate locations with which the plant has a need to communicate. This system bypasses external commercial telephone lines and switching equipment. Corporate transmission facilities provide fiber optic, copper-wire, and microwave radio to ensure a high degree of system reliability. In addition to the redundancy provided by the three system options, backup power is provided for the systems. DEMNET is the primary means of offsite communication. This circuit allows intercommunication among the EOF, TSC, control room, counties, and states. DEMNET operates as an internet based (VoIP)communications system with a satellite back-up. Should the internet transfer rate become slow or unavailable, the DEMNET will automatically transfer to satellite mode.OPEP-02.2.1 ev. 6 Page 226 of 310 ATTACHMENT 1 Page 201 of 219 EAL Bases Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by Duke Energy in four ways: (1) tie-ins through the PBX to any other plant location, (2) lines to plant emergency facilities, (3) lines to the Joint Information Center for public information purposes, and (4) lines to the AEF. The local service provider provides primary and secondary power for their lines at the Central Office.Satellite Phones A total of three portable satellite telephones are available which enable communication when all other phone systems are inoperable, e.g. following a major hurricane. These portable systems can be powered by internal batteries, external DC sources as well as external AC sources. Two of these phones require outside use, while one phone may used either outside or in the EOF with a permanently mounted external antenna.Cellular Phones Selected plant personnel are provided with cellular telephones. These phones may be used during emergencies if other communications means are not readily available or are inoperable. These phones are not expected to be used in the Control Room or Power Block due to interference with plant equipment and loss of signal to the phone.NRC Emergency Telecommunications System The NRC uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRC regional and national offices. The Duke Energy communications line provides a link independent of the local public telephone network.Telephones connected to this network are located in the Brunswick Control Room, Technical Support Center, and Emergency Operations Facility. These lines will not function if the PBX Telephone System fails.This EAL is the hot condition equivalent of the cold condition EAL CU5.1.NEI 99-01 Basis: This IC addresses a significant loss of oriite or offsite communications capabilities. While not -Deleted: -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 onpite -Deleted: -information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).I OPEP-02.2.1 I Rev. 6 1 Page 227 of 310 The first EAL condition addresses a total loss of the communications methods used in support -of routine plant operations. S-F Delted EAL #2.The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the State ------- -- D _(see De Brunswick and New Hanover County EOCs .--lDeleted:.,The third EAL addresses a total loss of the communications methods used to notifythe NRC Deleted: EAL#3 of an emergency declaration. ATTACHMENT 1 Page 202 of 219 EAL Bases I veoer Notes)BNP Basis Reference(s):

1. OERP Radiological Emergency Response Plan Appendix A 2. SD-48 Communication Systems 3. NEI 99-01 SU6 1OPEP-02.2.1 I Rev. 6 Page 228 of 310 ATTACHMENT 1 Page 203 of 219 EAL Bases Category: Subcategory:

Initiating Condition: S -System Malfunction 8 -Hazardous Event Affecting Safety Systems Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL: SA8.1 Alert The occurrence of any Table S-4 hazardous event AND EITHER:* 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 Table S-4 Hazardous Events" Seismic event (earthquake)" Internal or external FLOODING event" High winds or tornado strike" FIRE* EXPLOSION* Other events with similar hazard characteristics as determined by the Shift Manager Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown I OPEP-02.2.1 I Rev. 6 1 Page 229 of 310 ATTACHMENT 1 Page 204 of 219 EAL Bases Definition(s): 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 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.FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.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 (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. 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. BNP Basis:* The significance of seismic events are discussed under EAL HU2.1 (ref. 1).* Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 2, 3)." Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 135 mph. (ref. 4).OPEP-02.2.1 Rev. 6 Page 230 of 310 ATTACHMENT 1 Page 205 of 219 EAL Bases" Areas containing functions and systems required for safe shutdown of the plant are identified by Fire Zone in the fire response procedure (ref. 5, 6).* An explosion that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL.NEI 99-01 Basis: This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode.This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.The first condition addresses damage to a SAFETY SYSTEM train that is in service/operation -Deleted: EAL 1.b.1 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 condition addresses damage to a SAFETY SYSTEM component that is not in --Deleted: EAL 1.b.2 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 FS1 orRS1_ ------------ -- Deleted: ASi BNP Basis Reference(s):

1. 1(2)APP-UA-28 6-4 Seismic Event 2. OAOP-13.0 Operations During Hurricane, Flood Conditions, Tornado or Earthquake

3. Updated FSAR section 3.4.2 Protection From Internal Flooding 4. Updated FSAR Section 2.3.1.2.7 5. BNP-E-9.004 Safe Shutdown Analysis Report 6. OPFP-PBAA Power Block Auxiliary Areas Prefire Plan 7. NEI 99-01 SA9 IOPEP-02.2.1 I Rev. 6 Page 231 of 310 ATTACHMENT 1 Page 206 of 219 EAL Bases Category E -Independent Spent Fuel Storage Installation (ISFSI)EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold)An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a cask/canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel. Formal offsite planning is not required because the postulated worst-case accident involving an ISFSI has insignificant consequences to the public health and safety.An Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated.A hostile security event that leads to a potential loss in the level of safety of the ISFSI is a classifiable event under Security category EAL HS1.1.Minor surface damage that does not affect storage cask/canister boundary is excluded from the scope of these EALs.OPEP-02.2.1 Rev. 6 Page 232 of 310 ATTACHMENT 1 Page 207 of 219 EAL Bases Category:

E -ISFSI Sub-category: None Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL: EUI.1 Notification of Unusual Event Damage to a loaded canister confinement boundary as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask > any of the following:

1,400 mrem/hr on the HSM-H front surface* 10 mrem/hr on the HSM-H door centerline
20 mrem/hr on the end shield wall exterior Mode Applicability:

All Definition(s): CONFINEMENT BOUNDARY-. The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the BNP ISFSI, Confinement Boundary is defined as the Dry Shielded Canister (DSC).BNP Basis: The BNP ISFSI utilizes the NUHOMS Type 2 -61 BTH dry spent fuel storage (ref. 1, 2).The NUHOMS Type 2 61 BTH spent fuel storage system is a modular canister based spent fuel storage and transfer system and consists of the following components: " A 61BTH Dry Shielded Canister (DSC) provides confinement, an inert environment, structural support, and criticality control for 61 BWR fuel assemblies." A horizontal storage module (HSM-H) is provided for environmental protection, shielding, and heat rejection during storage.* An OS1 97FC-B transfer cask that supports onsite transfer of the 61BTH DSC.The NUHOMS System confinement vessel is the DSC. The DSC is welded and designed to provide confinement of all radionuclides under normal, off-normal, and accident conditions. IOPEP-02.2.1 I Rev. 6 Page 233 of 310 Confinement boundary is defined as the barrier(s) between areas containing radioactive substances and the environment. Therefore, damage to a confinement boundary must be a confirmed physical breach between the spent fuel and the environment for the Dry Shield Canister (DSC).The values shown represent 2 times the limits specified in the ISFSI Certificate of Compliance (COC) Technical Specification for radiation external to a loaded (NUHOMS Type 2 -61 BTH)MPC (HSM-H) overpack (ref. 1, 2). The survey method(s) used to assess this EAL threshold shall be consistent with those used to ensure compliance with the COC Technical Specification limits (ref. 2).NEI 99-01 Basis: This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", which is also used in Recognition Category,_ IC F_1, 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.-Deleted: A Deleted: A d Security-related events for ISFSIs are covered under ICs HU1 and HA1.OPEP-02.2.1 Rev. 1 Page 234 of 3101 ATTACHMENT 1 Page 209 of 219 EAL Bases BNP Basis Reference(s):

1. OPLP-36 BNP 1OCFR50.72.212 Report 2. Transnuclear, Inc. Standardized NUHOMS Horizontal Modular Storage System Certificate of Compliance No. 1004, Ammendment 10 Enclosure 1 3. NGGM-PM-0028 Transnuclear NUHOMS Dry Fuel Storage Program Manual 4. NEI 99-01 E-HU1 0PEP-02.2.1 ev. 6 Page 235 of 310 ATTACHMENT 1 Page 210 of 219 EAL Bases Category F -Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature

> 212°F); EALs in this category are applicable only in one or more hot operating modes.EALs in this category represent threats to the defense in depth design concept that precludes the release of highly 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: A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.B. Reactor Coolant System (RCS): 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.C. Containment (PC): The drywell, the torus, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves comprise the PC barrier. Primary Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency. The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1 (Attachment 2). "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials. "Potential Loss" means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level: Alert: Any loss or any potential loss of either Fuel Clad or RCS Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of third barrier I OPEP-02.2.1 I Rev. 6 1 Page 236 of 310 ATTACHMENT 1 Page 211 of 219 EAL Bases The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations: " The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Primary Containment Barrier." Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs." For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC RG1 has been exceeded.* The fission product barrier thresholds specified within a scheme reflect plant-specific BNP design and operating characteristics.

As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location-inside the primary containment, an interfacing system, or outside of the primary containment.

The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage.* At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the SEC would have more assurance that there was no immediate need to escalate to a General Emergency. OPEP-02.2.1 I Rev. 6 1 Page 237 of 310 I OPEP-02.2.1 Rev. 6 Page 238 of 310 1 ATTACHMENT 1 Page 213 of 219 EAL Bases Category: Subcategory: Fission Product Barrier Degradation N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS barrier EAL: FA1.1 Alert Any loss or any potential loss of either Fuel Clad or RCS (Table F-i)Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: Fuel Clad, RCS and Primary Containment comprise the fission product barriers. Table F-1 (Attachment

2) lists the fission product barrier thresholds, bases and references.

At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.1 NEI 99-01 Basis: None BNP Basis Reference(s):

1. NEI 99-01 FA1 I OPEP-02.2.1 Rev. 6 1 Page 239 of 310 ATTACHMENT 1 Page 214 of 219 EAL Bases Category: Subcategory:

Fission Product Barrier Degradation N/A Initiating Condition: Loss or potential loss of any two barriers EAL: FS1.1 Site Area Emergency Loss or potential loss of any two barriers (Table F-i)Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment

2) lists the fission product barrier thresholds, bases and references.

At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions: " One barrier loss and a second barrier loss (i.e., loss -loss)" One barrier loss and a second barrier potential loss (i.e., loss -potential loss)" One barrier potential loss and a second barrier potential loss (i.e., potential loss -potential loss)At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Site Emergency Coordinator would have greater assurance that escalation to a General Emergency is less imminent.IOPEP-02.2.1 I Rev. 6 Page 240 of 310 ATTACHMENT 1 Page 215 of 219 EAL Bases NEI 99-01 Basis: None BNP Basis Reference(s):

1. NEI 99-01 FS1 I OPEP-02.2.1 ev. 6 1Page 241 of 310 ATTACHMENT 1 Page 216 of 219 EAL Bases Category:

Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL: FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of third barrier (Table F-i)Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment

2) lists the fission product barrier thresholds, bases and references.

At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:

Loss of Fuel Clad, RCS and Containment barriers* Loss of Fuel Clad and RCS barriers with potential loss of Containment barrier* Loss of RCS and Containment barriers with potential loss of Fuel Clad barrier" Loss of Fuel Clad and Containment barriers with potential loss of RCS barrier NEI 99-01 Basis: None BNP Basis Reference(s):

1. NEI 99-01 FG1 IOPEP-02.2.1 I Rev. 6 Page 242 of 310 ATTACHMENT 2 Page 1 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Introduction Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment).

The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds. The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are: A. RPV Water Level B. RCS Leak Rate B. Primary Containment Conditions C. Primary Containment Radiation/RCS Activity D. Primary Containment Integrity or Bypass E. SEC Jugement Each category occupies a row in Table F-1 thus forming a matrix defined by the categories. The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell.Thresholds are assigned sequential numbers within each Loss and Potential Loss column beginning with number one. In this manner, a threshold can be identified by its category title and number. For example, the first Fuel Clad barrier Loss in Category A would be assigned"FC Loss A.1 ," the third Containment barrier Potential Loss would be assigned "PC P-Loss B.3," etc.If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers.When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-i, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been OPEP-02.2.1 I Rev. 6 1 Page 243 of 310 exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category I OPEP-02.2.1 I Rev. 6 1 Page 244 of 310 ATTACHMENT 2 Page 2 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost -even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS barriers and a Potential Loss of the Containment barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FS1.1, and FA1.1 to determine the appropriate emergency classification. In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the RCS barrier and finally the Containment barrier threshold bases. In each barrier, the bases are given according category Loss followed by category Potential Loss beginning with Category A, then B,..., F.I OPEP-02.2.1 I Rev. 6 1 Page 245 of 310 ATTACHMENT 2 Page 3 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier Reactor Coolant System Barrier Primary Containment Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A 1. RPV level cannot be restored and 1. RPV level cannot be restored and 1. Entry to SAMG-01 required maintained > TAF or cannot be maintained > TAF or cannot be None None 1. Entry to SAMG-01 required RPV Water determined determined Level 1. UNISOLABLE primary system 1. UNISOLABLE primary system leakage that results in exceeding leakage that resuots in exceeding 1. UNISOLABLE break in any of the EITHER of the following: EITHER of the following: tollowing: One or more Secondary

One or more Secondary SMain steam Containment area radiation Containment area radiation B HPCI steam Line Maedmum Normal Operating Matximum Sate Operating None None R tCIC steam Line Umits (SEOP-03-SCCP Table Limits (OEOP-03-SCCP Table None RCS Leak Rate
RWCU 3) 3)* Feedwater One or more Secondary
One or more Secondary 2. Emergency Dopressurization is Containment area temperature Containment area required Maximum Normal Operating temperature Madmom Safe ULmits Operating Limits (OEOP-03-SCCP Table 1) (0EOP-03-SCCP Table 1)1. UNPLANNED rapid drop in 1. Primary Containment pressure Primary Containment pressure >62 psig C following Primary Containment None None 1. Pnmary Containment pressure > None pressure rise 2. Deflagration concentrations eaist PC 1.7 psig due to RCS leakage inside PC (H 2 a 6% AND O2 a 50%)Condltions

2. Primary Containment pressure 3. Heat Capacity Temperature Limit response not consistent with (HCTL) exceeded LOCA conditions D 1. Drywall radiation

> 2.000 R/hr PC RadrI None 1. Drywell radiation >27 Fhnr with None None 1. Drywall radiation > 20.000WttRh RCS 2. Primary coolant activity > 300 reactot shutdown Activity pCi/gm 1-131 dose equivalent

1. UNISOLABLE direct downstream pathway to the environment E adsts atetr Primary Containment None None None None isolation signal None PC Integrity or Bypass 2. Intentional Primary Containment venting per EOPs F 1. Any condition in the opinion of 1. Any condition in the opinion of 1, AIy condition in the opinion of 1. Any condition in the opinion ol the 1. Aty condition in the opinion of 1. Any condition in the opinion of the the SEC that indicates loss of the SEC that indicates potential the SEC that indicates loss of the SEC that indicates potentiai loss of the SEC that Indicates loss of the SEC that indicates potential loss of ED the fuel dad bamer loss of the fuel clad barrier RCS barrier the RCS barner Primary Containment barrier the Primary Containment barner Judgmnent OPEP-02.2.1 Rev. 6 Page 246 of 310 ATTACHMENT 2 Page 4 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

A. RPV Level Degradation Threat: Loss Threshold:

1. Entry to SAMG-01 required Definition(s):

N/A BNP Basis: EOP-01-RVCP, EOP-01-LPC and EOP-01-RXFP specify the requirement for entry to SAMG-01 when core cooling is severely challenged. These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined. SAMG-01 is required when (ref. 1): " Reactor water level cannot be restored and maintained above -57.5 inches (Jet Pump Suction) with at least one core spray pump injecting into the reactor vessel" Reactor vessel water level cannot be restored and maintained above LL-4 (MSCRWL)" The reactor vessel flooding conditions cannot be restored and maintained (5 SRVs open and reactor vessel pressure more than 50 psig above suppression chamber pressure)" When at least 1 SRV cannot be opened and reactor vessel pressure cannot be restored and maintained above the minimum alternate reactor vessel flooding pressure (Table 1 values that are dependent on number of open SRVs)The above EOP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad.This threshold is also a Potential Loss of the Containment barrier (PC P-Loss A.1). Since entry to SAMG-01 occurs after core uncovery has occurred a Loss of the RCS barrier exists (RCS Loss A.1). Entry to SAMG-01, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification. ATTACHMENT 2 Page 5 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases OPEP-02.2.1 Rev. 6 Page 247 of 310 NEI 99-01 Basis: The Loss threshold represents the EOP requirement for entry to SAMG-01, This is identified- --Deleted: primary containment flooding in the BWROG EPGs/SAGs when the phrase, ".enter all Severe Accident Guidelines' appears. Formatted: Font: (Default) Arial Since a site-specific RPV water level is not specified here, the Loss threshold phrase, -Entry to Deleted: Primary Containment Flooding Is SAMG-01 required," also accommodates the EOP need to enter SAMG-01 when RPV water Required.level cannot be determined and core damage due to inadequate core cooling is believed to be Deleted: Primary containment flooding occurring. Deleted: flood the primary containment BNP Basis Reference(s):

1. OSAMG-06.0 SAMG Primary Containment Flooding Basis Document 2. NEI 99-01 RPV Water Level Fuel Clad Loss 2.A OPEP-02.2.1 Rev. 6 1 Page 248 of 310 ATTACHMENT 2 Page 6 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

A. RPV Level Degradation Threat: Potential Loss Threshold:

1. RPV level cannot be restored and maintained

> TAF or cannot be determined Definition(s): N/A BNP Basis: An RPV level instrument reading of -7.5 in. indicates RPV level is at the top of active fuel (TAF) (ref. 1). When RPV level is at or above the TAF, the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV level is below TAF, the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray cooling). If core uncovery is threatened, the EOPs specify alternate, more extreme, RPV level control measures in order to restore and maintain adequate core cooling.Since core uncovery begins if RPV level drops to TAF, the level is indicative of a challenge to core cooling and the Fuel Clad barrier.When RPV level cannot be determined, EOPs require entry to EOP-01 -RXFP, Reactor Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained (ref. 2, 3). When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EOP-01 -RXFP specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Alternate Reactor Vessel Flooding Pressure (in scram-failure events) (ref. 4). If RPV water level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists.Note that EOP-01 -LPC, Level/Power Control, may require intentionally lowering RPV water level to TAF and control level between the LL-4, the Minimum Steam Cooling RPV Water Level (MSCRWL) and TAF (ref. 3). Under these conditions, a high-power ATWS event exists and requires at least an Alert classification in accordance with the System Malfunction -RPS Failure EALs, however under these conditions a potential loss of the fuel clad does not exist.IOPEP-02.2.1 I Rev. 6 Page 249 of 310 ATTACHMENT 2 Page 7 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases NEI 99-01 Basis: 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 1.A. Thus, this --Deleted: 2 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 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. 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 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 SS_,wil/ dictate the need for emergency classification. -- -Dele" ,Deleted: 5J 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. I OPEP-02.2.1 I Rev. 6 Page 250 of 310 1 ATTACHMENT 2 Page 7 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases BNP Basis Reference(s):

1. OEOP-01-NL EOP-SAMG Numerical Limits and Values, Attachment 1 2. EOP-01-RVCP Reactor Vessel Control 3. EOP-01 -LPC, Level/Power Control 4. EOP-01-RXFP, Reactor Flooding 5. NEI 99-01 RPV Water Level Fuel Clad Potential Loss 2.A OPEP-02.2.1 ev. 6 Page 25 of30 ATTACHMENT 2 Page 8 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

B. RCS Leak Rate Degradation Threat: Loss Threshold: None i OPEP-02.2.1 I Rev. 6 Page 252 of 310 ATTACHMENT 2 Page 9 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. RCS Leak Rate Degradation Threat: Potential Loss Threshold: None I OPEP-02.2.1 I Rev. 6 Page 253 of 310 1 ATTACHMENT 2 Page 10 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. PC Conditions Degradation Threat: Loss Threshold: None I OPEP-02.2.1 Rev. 6 1 Page 254 of 310 ATTACHMENT 2 Page 11 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. PC Conditions Degradation Threat: Potential Loss Threshold: None IOPEP-02.2.1 I Rev. 6 1 Page 255 of 310 1 ATTACHMENT 2 Page 12 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. PC Radiation / RCS Activity Degradation Threat: Loss Threshold:

1. Drywell radiation

> 2,000 R/hr Definition(s): None BNP Basis: The Drywell High-Range Radiation Monitor (1 (2)D22-RI-4195, 1 (2)D22-RI-4196, 1 (2)D22-RI-4197, 1 (2)D22-RI-4198) reading of 2,000 R/hr indicates the release of reactor coolant into the drywell with elevated activity indicative of fuel damage.Based on 2% clad damage, a containment radiation level of 2000 R/hr is derived as follows: Per OPEP-03.6.3 Table 3, 100% Cladding Damage column 'No Spray' for 1 hour after shutdown is 100,000 R/hr. Per Step 7.2.2.1, 0.02 x 100,000 R/hr = 2000 R/hr containment radiation (ref. 1).NEI 99-01 Basis: 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 l.Ci/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage.Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold D.1 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. IOPEP-02.2.1 I Rev. 6 Page 256 of 310 ATTACHMENT 2 Page 12 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases BNP Basis Reference(s):

1. OPEP-03.6.3 Estimate of the Extent of Core Damage Under Accident Conditions

2. NEI 99-01 Primary Containment Radiation Fuel Clad Loss 4.A OPEP-02.2.1 ev. 6 Page 257 of 310 ATTACHMENT 2 Page 13 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

D. PC Radiation / RCS Activity Degradation Threat: Loss Threshold:

2. Primary coolant activity > 300 pCi/gm 1-131 dose equivalent BNP Basis: None NEI 99-01 Basis: This threshold indicates that RCS radioactivity concentration is greater than 300 pCi/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.There is no Potential Loss threshold associated with Primary Containment Radiation.

BNP Basis Reference(s):

1. NEI 99-01 RCS Activity Fuel Clad Loss 1.A I OPEP-02.2.1 Rev. 6 1 Page 258 of 310 ATTACHMENT 2 Page 14 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

D. PC Radiation / RCS Activity Degradation Threat: Potential Loss Threshold: None I OPEP-02.2.1 I Rev. 6 1 Page 259 of 310 ATTACHMENT 2 Page 14 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. PC Integrity or Bypass Degradation Threat: Loss Threshold: None I OPEP-02.2.1 I Rev. 6 1 Page 260 of 310 ATTACHMENT 2 Page 14 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. PC Integrity or Bypass Degradation Threat: Potential Loss Threshold: None I I OPEP-02.2.1 I Rev. 6 1 Page 261 of 310 ATTACHMENT 2 Page 15 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: F. SEC Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Site Emergency Coordinator that indicates loss of the Fuel Clad barrier Definition(s):

None BNP Basis: The Site Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Site Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that are to be used by the Site Emergency Coordinator in determining whether the Fuel Clad barrier is lost ----- Deleted: Director BNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A OPEP-02.2.1 IRev. 6 1Page 262 of 310 ATTACHMENT 2 Page 16 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

F. SEC Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Site Emergency Coordinator that indicates potential loss of the Fuel Clad barrier BNP Basis: The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences." Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Site Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that are to be used by the Site Emergency Coordinator-in determining whether the Fuel Clad barrier is potentially lost. The Site --F Deleted- Director Emergency Coordinator-should also consilder whet-h-er or not to d eclare- the -barrierpotentially_ .. -Deleted: Director lost in the event that barrier status cannot be monitored. BNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A OPEP-02.2.1 Rev. 6 Page 263 of 310 ATTACHMENT 2 Page 17 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

A. RPV Water Level Degradation Threat: Loss Threshold:

1. RPV level cannot be restored and maintained

> TAF or cannot be determined Definition(s): None BNP Basis: An RPV level instrument reading of -7.5 in. indicates RPV level is at the top of active fuel (TAF) (ref. 1). TAF is significantly lower than the normal operating RPV level control band. To reach this level, RPV inventory loss would have previously required isolation of the RCS and Primary Containment barriers, and initiation of all ECCS. If RPV level cannot be maintained above TAF, ECCS and other sources of RPV injection have been ineffective or incapable of reversing the decreasing level trend. The cause of the loss of RPV inventory is therefore assumed to be a LOCA. By definition, a LOCA event is a Loss of the RCS barrier.When RPV level cannot be determined, EOPs require entry to EOP-01 -RXFP, Reactor Flooding (ref. 2). The instructions in EOP-01 -RXFP specify emergency depressurization of the RPV, which is defined to be a Loss of the RCS barrier (RCS Loss C.4).The conditions of this threshold are also a Potential Loss of the Fuel Clad barrier (FC P-Loss A.1). A Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier requires a Site Area Emergency classification. Note that EOP-01-LPC, Level/Power Control, may require intentionally lowering RPV water level to TAF and control level between LL-4, the Minimum Steam Cooling RPV Water Level (MSCRWL), and TAF (ref. 3). Under these conditions, a high-power ATWS event exists and requires at least an Alert classification in accordance with the System Malfunction -RPS Failure EALs, however under these conditions a loss of the RCS does not exist.OPEP-02.2.1 Rev. 6 Page 264 of 310 ATTACHMENT 2 Page 18 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases NEI 99-01 Basis: 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 .A. _ --Deleted: 2 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 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. BNP Basis Reference(s):

1. OEOP-01 -NL EOP-SAMG Numerical Limits and Values, Attachment 1 2. EOP-01-RVCP Reactor Vessel Control 3. EOP-01-LPC, Level/Power Control 4. EOP-01-RXFP, Reactor Flooding 5. NEI 99-01 RPV Water Level RCS Loss 2.A I OPEP-02.2.1 I Rev. 6 1 Page 265 of 310 ATTACHMENT 2 Page 19 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

A. RPV Water Level Degradation Threat: Potential Loss Threshold: None IOPEP-02.2.1 I Rev. 6 1 Page 266 of 310 ATTACHMENT 2 Page 22 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:

1. UNISOLABLE break outside Primary Containment in any of the following: " Main steam line* HPCI steam line* RCIC steam line* RWCU* Feedwater Definition(s):

UNISOLABLE-An open or breached system line that cannot be isolated, remotely or locally.BNP Basis: The conditions of this threshold include required containment isolation failures allowing a flow path to the environment. A release pathway outside primary containment exists when flow is not prevented by downstream isolations. In the case of a failure of both isolation valves to close but in which no downstream flowpath exists, emergency declaration under this threshold would not be required. Similarly, if the emergency response requires the normal process flow of a system outside primary containment (e.g., EOP requirement to bypass MSIV low RPV water level interlocks and maintain the main condenser as a heat sink using main turbine bypass valves), the threshold is not met. The combination of these threshold conditions represent the loss of both the RCS and Primary Containment (see PC Loss E.1) barriers and justifies declaration of a Site Area Emergency (i.e., Loss or Potential Loss of any two barriers). Even though RWCU and Feedwater systems do not contain steam, they are included in the list because an unisolable break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume systems directly connected to the RCS.NEI 99-01 Basis: 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.IOPEP-02.2.l Rev. 6 1Page 267 of 310 ATTACHMENT 2 Page 23 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases BNP Basis Reference(s):

1. 1(2)OP-01 Nuclear Boiler System 2. 1(2)OP-25 Main Steam System Operating Procedure 3. 1(2)OP-19 High Pressure Coolant Injection System Operating Procedure 4. 1(2)OP-1 6 Reactor Core Isolation Cooling System Operating Procedure 5. 1(2)OP-1 4 Reactor Water Cleanup System Operating Procedure 6. 1(2)OP-32 Condensate and Feedwater System Operating Procedure 7. NEI 99-01 RCS Leak Rate RCS Loss 3.A IOPEP-02.2.1 I Rev. 6 1 Page 268 of 310 ATTACHMENT 2 Page 24 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

B. RCS Leak Rate Degradation Threat: Loss Threshold:

2. Emergency Depressurization is required Definition(s):

N/A BNP Basis: Plant symptoms requiring Emergency Depressurization per the EOPs are indicative of a loss of the RCS barrier. If Emergency depressurization is required, the plant operators are directed to open safety relief valves (SRVs) and keep them open regardless of any subsequent radiological release rate (ref. 1 -6). Even though the RCS is being vented into the suppression pool, a loss of the RCS should be considered to exist due to the diminished effectiveness of the RCS pressure barrier to a release of fission products beyond its boundary.NEI 99-01 Basis: 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) 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.BNP Basis Reference(s):

1. EOP-01-UG User's Guide 2. EOP-01-RVCP Reactor Vessel Control 3. EOP-01 -LCP Level/Power Control 4. EOP-02-PCCP Primary Containment Control 5. OEOP-03-SCCP Secondary Containment Control 6. EOP-04-RRCP Radioactivity Release Control 7. EOP-01-RXFP Reactor Flooding 8. NEI 99-01 RCS Leak Rate RCS Loss 3.B OPEP-02.2.1 IRev. 6 Page 269 of 310 ATTACHMENT 2 Page 26 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Reactor Coolant System B. RCS Leak Rate Barrier: Category: Degradation Threat: Potential Loss Threshold:

1. UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

One or more Secondary Containment area radiation Maximum Normal Operating Limits (OEOP-03-SCCP Table 3)" One or more Secondary Containment area temperature Maximum Normal Operating Limits (OEOP-03-SCCP Table 1)Definition(s):

UNISOLABLE-An open or breached system line that cannot be isolated, remotely or locally.BNP Basis: The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of unisolable primary system leakage outside the primary containment. The Maximum Normal Operating Limit values define this RCS threshold because they are the maximum normal operating values and signify the onset of abnormal system operation. When parameters reach this level, equipment failure or misoperation may be occurring. Elevated parameters may also adversely affect the ability to gain access to or operate equipment within the affected area. The locations into which the primary system discharge is of concern correspond to the areas addressed in OEOP-03-SCCP, Secondary Containment Control Tables (ref. 1).In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g.room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment. IOPEP-02.2.1 I Rev. 6 1 Page 270 of 310 ATTACHMENT 2 Page 27 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases NEI 99-01 Basis: 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.BNP Basis Reference(s):

1. OEOP-03-SCCP, Secondary Containment Control 2. NEI 99-01 RCS Leak Rate RCS Potential Loss 3.A IOPEP-02.2.1 ev. 6 Page 271 of 310 ATTACHMENT 2 Page 20 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

C. PC Conditions Degradation Threat: Loss Threshold:

1. Primary Containment pressure > 1.7 psig due to RCS leakage Definition(s):

None BNP Basis: The drywell high pressure scram setpoint is an entry condition to EOP-01 -RVCP Reactor Vessel Control, and EOP-02-PCCP, Primary Containment Control (ref. 1, 2, 3). Normal primary containment pressure control functions (e.g., operation of drywell coolers, vent through SBGT, etc.) are specified in EOP-02-PCCP in advance of less desirable but more effective functions (e.g., operation of drywell or suppression pool sprays, etc.).In the BNP design basis, primary containment pressures above the drywell high pressure scram setpoint are assumed to be the result of a high-energy release into the containment for which normal pressure control systems are inadequate or incapable of reversing the increasing pressure trend. Pressures of this magnitude, however, can be caused by non-LOCA events such as a loss of drywell cooling or inability to control primary containment vent/purge (ref. 4).The threshold phrase "...due to RCS leakage" focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. PC pressure greater than 1.7 psig with corollary indications (e.g., drywell temperature, indications of loss of RCS inventory) should therefore be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressure greater than 1.7 psig should not be considered an RCS barrier Loss.NEI 99-01 Basis: 1.7 ic-- Deleted: The (site-specific value) primary 17s- isthedrywell high pressure setpoint which indicates a LOCA by automatically----- I containment pressure initiating the ECCS or equivalent makeup system.There is no Potential Loss threshold associated with Primary Containment Pressure.OPEP-02.2.1 Rev. 6 Page 272 of 310 ATTACHMENT 2 Page 20 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases BNP Basis Reference(s):

1. OEOP-01-NL EOP-SAMG Numerical Limits and Values, Attachment 3 2. EOP-01-RVCP Reactor Vessel Control 3. EOP-02-PCCP Primary Containment Control 4. BNP Updated FSAR Chapter 6 Emergency Core Cooling Systems 5. NEI 99-01 Primary Containment Pressure RCS Loss 1.A I OPEP-02.2.1 I Rev. 6 1 Page 273 of 310 ATTACHMENT 2 Page 21 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

C. PC Conditions Degradation Threat: Potential Loss Threshold: None I OPEP-02.2.1 Rev. 6 1 Page 274 of 310 ATTACHMENT 2 Page 29 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. PC Radiation / RCS Activity Degradation Threat: Loss Threshold:

1. Drywell radiation

> 27 R/hr with reactor shutdown Definition(s): N/A BNP Basis: The Drywell High-Range Radiation Monitor (1 (2)D22-RI-4195, 1 (2)D22-RI-4196, 1 (2)D22-RI-4197, 1 (2)D22-RI-4198) reading of 27 R/hr is based on coolant activity at the Technical Specification limit of 4 pCi/gm 1-131).The containment radiation level of 27 R/hr is derived as follows: 0PEP-03.6.3 Table 3 100% Cladding Damage column 'No Spray' for 1 hour after shutdown is 100,000 R/hr. Assuming that 300 pCi/gm 1-131 is approximately 2% cladding failure, a coolant activity of 4 pICi/gm 1-131 is ratioed to approximately 0.027% (0.00027) clad failure. Per Step 7.2.2.1, 0.00027 x 100,000 R/hr = 27 R/hr containment radiation corresponding to Technical Specification coolant activity. (ref. 1)The threshold value is only applicable with the reactor shutdown as the high range detectors normally read as high as 100 R/hr during power operations due to shine from the reactor.NEI 99-01 Basis: 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 D.1 since it indicates a loss of the RCS Barrier only.There is no Potential Loss threshold associated with Primary Containment Radiation. IOPEP-02.2.1 I Rev. 6 1 Page 275 of 310 ATTACHMENT 2 Page 29 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases BNP Basis Reference(s):

1. OPEP-03.6.3 Estimate of the Extent of Core Damage Under Accident Conditions

2. NEI 99-01 Primary Containment Radiation RCS Loss 4.A IOPEP-02.2.1 IRev. 6 1Page 276 of 310 ATTACHMENT 2 Page 30 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

D. PC Radiation / RCS Activity Degradation Threat: Potential Loss Threshold: None I OPEP-02.2.1 Rev. 6 1 Page 277 of 310 ATTACHMENT 2 Page 30 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. PC Integrity or Bypass Degradation Threat: Loss Threshold: None I OPEP-02.2.1 Rev. 6 1 Page 278 of 310 ATTACHMENT 2 Page 30 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. PC Integrity or Bypass Degradation Threat: Potential Loss Threshold: None I OPEP-02.2.1 Rev. 6 1 Page 279 of 310 ATTACHMENT 2 Page 31 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: F. SEC Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Site Emergency Coordinator that indicates loss of the RCS barrier Definition(s):

None BNP Basis: The Site Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent barrier deqradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to the recognition of the inability to reach safety acceptance criteria before completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Site Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that may be used by the Site Emergency Coordinator in determining whether the RCS Barrier is lost. --- Deleted: Director BNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A I OPEP-02.2.1 Rev. 6 1 Page 280 of 310 ATTACHMENT 2 Page 32 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

F. SEC Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Site Emergency Coordinator that indicates potential loss of the RCS barrier Definition(s):

None BNP Basis: The Site Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to the inability to reach final safety acceptance criteria before completing all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Site Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that may be used by the Site Emergency Coordinator in determining whether the RCS Barrier is ptotiallyJost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. BNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A OPEP-02.2.1 Rev. 6 of 101---Deleted: Director ATTACHMENT 2 Page 33 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

A. RPV Water Level Degradation Threat: Loss Threshold: None I OPEP-02.2.1 Rev. 6 1 Page 282 of 310 ATTACHMENT 2 Page 34 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category: A. RPV Water Level Degradation Threat: Potential Loss Threshold:

1. Entry to SAMG-01 required Definition(s):

None BNP Basis: EOP-01 -RVCP, EOP-01 -LPC and EOP-01 -RXFP specify the requirement for entry to SAMG-01 when core cooling is severely challenged. These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined. Entry to SAMG-01 is required when (ref. 1):* Reactor water level cannot be restored and maintained above -57.5 inches (Jet Pump Suction) with at least one core spray pump injecting into the reactor vessel" Reactor vessel water level cannot be restored and maintained above LL-4 (MSCRWL)* The reactor vessel flooding conditions cannot be restored and maintained (5 SRVs open and reactor vessel pressure more than 50 psig above suppression chamber pressure)" When at least 1 SRV cannot be opened and reactor vessel pressure cannot be restored and maintained above the minimum alternate reactor vessel flooding pressure (EOP-01-RXFP Table 1 values that are dependent on number of open SRVs)The above EOP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad.This threshold is also a Loss of the Fuel Clad barrier (FC Loss A.1). Since entry to SAMG-01 occurs after core uncovery has occurred a Loss of the RCS barrier exists (RCS Loss A.1).Entry to SAMG-01, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification. OPEP-02.2.1 I Rev. 6 Page 283 of 310 ATTACHMENT 2 Page 35 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases NEI 99-01 Basis: The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold.A.1 -Deleted: 2.A The Potential Loss requirement for entry to SAMG-01 indicates adequategcore coolingcannot --Deleted: Primary Containment Flooding be restored and maintained and that core damage is possible. BWR EPGs/SAGs specify the conditions that require entry to SAMG-01,. Wheln_entry to SAMG-01 is required, the EPGs are --Deleted: primary containment flooding exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the -Deleted: primary containment flooding inability to restore and maintain 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. BNP Basis Reference(s):

1. OSAMG-06.0 SAMG Primary Containment Flooding Basis Document 2. NEI 99-01 RPV Water Level PC Potential Loss 2.A OPEP-02.2.1 Rev. 6 Page 284 of 310 ATTACHMENT 2 Page 46 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

B. RCS Leak Rate Degradation Threat: Loss Threshold:

1. UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

One or more Secondary Containment area radiation Maximum Safe Operating Limits (OEOP-03-SCCP Table 3)" One or more Secondary Containment area temperature Maximum Safe Operating Limits (OEOP-03-SCCP Table 1)Definition(s):

UNISOLABLE-An open or breached system line that cannot be isolated, remotely or locally.BNP Basis: The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of unisolable primary system leakage outside the primary containment. The Maximum Safe Operating Limit values define this Containment barrier threshold because they are indicative of problems in the secondary containment that are spreading and pose a threat to achieving a safe plant shutdown. This threshold addresses problematic discharges outside primary containment that may not originate from a high-energy line break. The locations into which the primary system discharge is of concern correspond to the areas addressed in OEOP-03-SCCP, Secondary Containment Control Tables 3 and 1 (ref.1) (see below). It should be noted that the Maximum Safe Radiation Operating Limits generally cannot be read in the Control Room and require local survey to assess.In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g.room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment. IOPEP-02.2.1 I Rev. 6 Page 285 of 310 ATTACHMENT 2 Page 47 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases NEI 99-01 Basis: 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,2.A this threshold would result in a Site-Area Area---Emergency. There is no Potential Loss threshold associated with Primary Containment Isolation Failure.BNP Basis Reference(s):

1. OEOP-03-SCCP Secondary Containment Control 2. NEI 99-01 RCS Leak Rate PC Loss 3.C-A Deleted: 3 IOPEP-02.2.1 Rev. 6 Page 286 of 3101 ATTACHMENT 2 Page 46 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

B. RCS Leak Rate Degradation Threat: Potential Loss Threshold: None I OPEP-02.2.1 R Rev. 6 Page 287 of 310 ATTACHMENT 2 Page 36 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category: C. PC Conditions Degradation Threat: Loss Threshold:

1. UNPLANNED rapid drop in Primary Containment pressure following Primary Containment pressure rise Definition(s):

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.BNP Basis: None NEI 99-01 Basis: 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 ................................................... , This thresholdelies 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. BNP Basis Reference(s):

1. NEI 99-01 Primary Containment Conditions PC Loss 1.A---Deleted: Primary containment pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure notI increasing under these conditions indicates a loss of primary containment integrity.

Deleted: These Deleted: a Deleted: rely I OPEP-02.2.1 Rev. 6 1 Page 288 of 310 ATTACHMENT 2 Page 37 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category: C. PC Conditions Degradation Threat: Loss Threshold:

2. Primary Containment pressure response not consistent with LOCA conditions Definition(s):

None BNP Basis: The calculated pressure response of the containment is shown in Figure 6-11. Figure 6-11 shows that the maximum calculated drywell pressure is 48 psia (33 psig), which is well below the design allowable pressure of 62 psig (ref. 2). The primary containment pressure stabilizes at about 40 psia (25 psig), as shown on Figure 6-1.Due to conservatisms in LOCA analyses, actual pressure response is expected to be less than the analyzed response. For example, blowdown mass flowrate may be only 60-80% of the analyzed rate.NEI 99-01 Basis: I Deleted: Rapid UNPLANNED loss of primary Primary containment pressure should increase as a result of mass and energy release into the .Deleted: pid UPAeD los aftrimary-- -- ---- --I containment pressure (i.e., not attributable to drywell spray or condensation effects) following primary containment from a LOCA. Thus, primary containment pressure not increasing under / an initial pressure increase indicates a loss of primary containment integrity. these conditions indicates a loss of primary containment integrity. This thresholdj.elies on operator recognition of an unexpected response for the condition and ---ted: These therefore a specific value is not assigned. The unexpected (UNPLANNED) response is I Deleted: important because it is the indicator for a containment bypass condition. BNP Basis Reference(s):

1. BNP Updated FSAR Figure 6-11 OPEP-02.2.1 I Rev. 6 Page 289 of 310

2. BNP Updated FSAR section 6.2.1.1.1 3. NEI 99-01 Primary Containment Conditions PC Loss 1.8 IOPEP-02.2.1 ev. 1 Page 290 of 310 ATTACHMENT 2 Page 38 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases UPDATED FSAR Revisim: I&A CP&L E.,'GrREmDS.4-nTF,4nRMs Figure: 6-11 tftW-E-p -. CIPer. 6 FIG-TRE Page: 1 ofl CONTAINMENT PRESSURE RESPONSE TO DESIGN BASIS ACCIDENT LOCA AT 120 PERCENT UPRATE 50 40 30-10 100 1000 M000D 100000 1000000 Time (seconfda OPEP-02.2.1 Rev. 6 Page 291 of 310 ATTACHMENT 2 Page 39 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

B. PC Conditions Degradation Threat: Potential Loss Threshold:

1. Primary Containment pressure > 62 psig Definition(s):

None BNP Basis: When the primary containment exceeds the maximum allowable value (62 psig) (ref. 1), primary containment venting may be required even if offsite radioactivity release rate limits will be exceeded (ref. 2). The drywell and suppression chamber maximum allowable value of 62 psig is based on the primary containment design pressure as identified in the BNP accident analysis (ref. 1, 3). If this threshold is exceeded, a challenge to the containment structure has occurred because assumptions used in the accident analysis are no longer valid and an unanalyzed condition exists. This constitutes a Potential Loss of the Containment barrier even if a containment breach has not occurred.NEI 99-01 Basis: 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.BNP Basis Reference(s):

1. BNP Updated FSAR section 6.2.1.1.1 2. EOP-02-PCCP Primary Containment Control 3. BNP Updated FSAR section 6.2.1.1 4. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.A OPEP-02.2.1 IRev. 6 1Page 292 of 310 ATTACHMENT 2 Page 40 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

B. PC Conditions Degradation Threat: Potential Loss Threshold:

2. Deflagration concentrations exist inside PC (H 2 > 6% AND 02 > 5%)Definition(s):

None BNP Basis: Deflagration (explosive) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAMGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (ref. 1).Except for brief periods during plant startup and shutdown, oxygen concentration in the primary containment is maintained at insignificant levels by nitrogen inertion. The specified values for this Potential Loss threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, ref. 2) and readily recognizable because 6% hydrogen is well above the EOP-02-PCCP, Primary Containment Control, entry condition (ref. 2). The minimum global deflagration hydrogen/oxygen concentrations (6% and 5%, respectively) require intentional primary containment venting, which is defined to be a Loss of Containment (PC Loss C.4).Monitors CAC-AT-4409 and 4410 monitor hydrogen and oxygen concentration in the drywell during normal operation. These monitors are isolated by accident isolation signals. However, monitors CAC-AT-4409 and 4410 will be realigned to the primary containment for post-accident monitoring via an operator actuated isolation signal override circuit when directed by the EOPs.oPEP-02.2.1 Rev. 6 Page 293 of 310 ATTACHMENT 2 Page 41 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases The oxygen and hydrogen concentrations from these two analyzers are recorded on two 4-channel recorders (CAC-AR-4409 and 4410) located on Panel XU-51. The indications are also displayed on the ERFIS. If concentrations exceed preset levels, recorder CAC-AR-4409 will annunciate the "Containment Atmosphere Division I 02 -H2 High" alarm in the Control Room and recorder CAC-AR-441 0 will annunciate "Containment Atmosphere Division II 02 -H2 High" alarm. (ref. 3, 4).NEI 99-01 Basis: 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. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur.BNP Basis Reference(s):

1. BWROG EPG/SAG Revision 2, Sections PC/G 2.3.EOP-02-PCCP, Primary Containment Control BNP Updated FSAR section 6.2.5.2.2 4. BNP System Description SD-04 Primary Containment

5. NEI 99-01 Primary Containment Conditions PC Potential Loss 1..B OPEP-02.2.1 Rev. 1 Page 294 of 310 ATTACHMENT 2 Page 42 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

B. PC Conditions Degradation Threat: Potential Loss Threshold:

3. Heat Capacity Temperature Limit (HCTL) exceeded Definition(s):

None BNP Basis: This threshold is met when the final step of section SP/T in EOP-02-PCCP, Primary Containment Control, is reached (ref. 1, 2).NEI 99-01 Basis: The Heat Capacity Temperature Limit (HCTL) 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 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 HCTL 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 OPEP-02.2.1 I Rev. 6 1 Page 295 of 310 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. BNP Basis Reference(s):

1. EOP-01-NL EOP/SAMG Numerical Limits and Values 2. EOP-02-PCCP Primary Containment Control 3. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.C IOPEP-02.2.1 I Rev. 6 1 Page 296 of 3101 ATTACHMENT 2 Page 49 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

D. PC Radiation / RCS Activity Degradation Threat: Loss Threshold: None I OPEP-02.2.1 I Rev. 6 1 Page 297 of 310 ATTACHMENT 2 Page 50 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category: D. PC Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:

1. Drywell radiation

> 20,000 R/hr Definition(s): None BNP Basis: The Drywell High-Range Radiation Monitor (1 (2)D22-RI-4195, 1 (2)D22-RI-4196, 1 (2)D22-RI-4197, 1(2)D22-RI-4198) reading of 20,000 R/hr indicates the release of reactor coolant into the drywell with elevated activity indicative of fuel damage.Based on 20% clad damage, a containment radiation level of 20,000 R/hr is derived as follows: 0PEP-03.6.3 Table 3 100% Cladding Damage column 'No Spray' for 1 hour after shutdown is 100,000 R/hr. Per Step 7.2.2.1, 0.2 x 100,000 R/hr = 20,000 R/hr containment radiation corresponding to 20% clad damage.In order to reach this Containment barrier Potential Loss threshold, a loss of the RCS barrier (RCS Loss D.5) and a loss of the Fuel Clad barrier (FC Loss D.2) have already occurred. This threshold, therefore, represents at a General Emergency classification. NEI 99-01 Basis: 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. 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. OPEP-02.2.1 IRev. 6 1Page 298 of 310 ATTACHMENT 2 Page 50 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases BNP Basis Reference(s):

1. QPEP-03.6.3 Estimate of the Extent of Core Damage Under Accident Conditions I OPEP-02.2.1 I Rev. 6 1 Page 299 of 310 ATTACHMENT 2 Page 43 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

E. PC Integrity or Bypass Degradation Threat: Loss Threshold:

1. UNISOLABLE direct downstream pathway to the environment exists after Primary Containment isolation signal Definition(s):

UNISOLABLE-An open or breached system line that cannot be isolated, remotely or locally.BNP Basis: This threshold addresses failure of open isolation devices which should close upon receipt of a manual or automatic containment isolation signal resulting in a significant radiological release pathway directly to the environment. The concern is the unisolable open pathway to the environment. A failure of the ability to isolate any one line indicates a breach of primary containment integrity. As stated above, the adjective "Direct" modifies "release pathway" to discriminate against release paths through interfacing liquid systems. Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include unisolable Main steam line, HPCI steam line or RCIC steam line breaks, unisolable RWCU system breaks, and unisloable containment atmosphere vent paths. If the main condenser is available with an unisolable main steam line, there may be releases through the steam jet air ejectors and gland seal exhausters. These pathways are monitored, however, and do not meet the intent of a nonisolable release path to the environment. These minor releases are assessed using the Category R, Abnormal Rad Release / Rad Effluent, EALs.The existence of an in-line charcoal filter (SBGT) does not make a release path indirect since the filter is not effective at removing fission noble gases. Typical filters have an efficiency of 95-99% removal of iodine. Given the magnitude of the core inventory of iodine, significant releases could still occur. In addition, since the fission product release would be driven by boiling in the reactor vessel, the high humidity in the release stream can be expected to render the filters ineffective in a short period.The threshold is met if the breach is not isolable from the Control Room or an attempt for isolation from the Control Room has been made and was unsuccessful. An attempt for isolation from the Control Room should be made prior to the emergency classification. If operator actions from the Control Room are successful, this threshold is not applicable. Credit OPEP-02.2.1 I Rev. 6 1 Page 300 of 310 is not given for operator actions taken in-plant (outside the Control Room) to isolate the breach.EOP-02-PCCP, Primary Containment Control, Section PC/P may specify primary containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1). Under these conditions with a valid containment isolation signal, the Containment barrier should be considered lost.NEI 99-01 Basis: 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 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 ICs ............................................ I -Deleted: A I BNP Basis Reference(s):

1. EOP-02-PCCP Primary Containment Control 2. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.A iOPEP-02.2.1 ev. 6 Page 301 of 31 ATTACHMENT 2 Page 45 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

E. PC Integrity or Bypass Degradation Threat: Loss Threshold:

2. Intentional Primary Containment venting per EOPs Definition(s):

None BNP Basis: EOP-02-PCCP, Primary Containment Control, may specify primary containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1, 2). The threshold is met when the operator begins venting the primary containment in accordance with EOP-01-SEP-01, not when actions are taken to bypass interlocks prior to opening the vent valves. Purge and vent actions specified in step PC/P-03 to control drywell pressure below the drywell high pressure scram setpoint or in section PC/H does not meet this threshold because such action is only permitted if offsite radioactivity release rates will remain below the ODCM limits.NEI 99-01 Basis: 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 control to the -combustible gas secondary containment and/or the environment is a Loss of the Containment. 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. BNP Basis Reference(s):

1. EOP-02-PCCP Primary Containment Control 2. EOP-01-SEP-01 Primary Containment Venting 3. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.B OPEP-02.2.1 Rev. 6 1 Page 302 of 310 ATTACHMENT 2 Page 49 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

E. PC Integrity or Bypass Degradation Threat: Potential Loss Threshold: None I OPEP-02.2.1 I Rev. 6 1 Page 303 of 310 ATTACHMENT 2 Page 51 of 54 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category: F. SEC Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Site Emergency Coordinator that indicates loss of the Primary Containment barrier Definition(s):

None BNP Basis: The Site Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Site Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that may be used by the Site Emergency Coordinator in determining whether the Primary Containment Barrier is lost. ----- Deleted: Director BNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A IOPEP-02.2.1 I Rev. 6 Page 304 of 310 ATTACHMENT 2 Page 53 of 53 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

F. SEC Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Site Emergency Coordinator that indicates potential loss of the Primary Containment barrier Definition(s):

None BNP Basis: The Site Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Site Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that may be used by the Site Emergency Coordinator in determining whether the Primary Containment Barrier is lost. ................ --Deleted: Director BNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A I OPEP-02.2.1 Rev. 6 1 Page 305 of 310 ATTACHMENT 3 Page 1 of 4 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases Background NEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located.These areas are intended to be plant operating mode dependent. Specifically the Developers Notes For AA3 and HA5 states: The "site-specific list of plant rooms or areas with entry-related mode applicability identified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area.The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections). Further, as specified in IC HA5: The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas.Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope.I OPEP-02.2.1 I Rev. 6 1 Page 306 of 310 ATTACHMENT 3 Page 2 of 4 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases BNP Table R-2 and H-2 Bases A review of station operating procedures identified the following mode dependent in-plant actions and associated safe shutdown areas that are required for normal plant operation, cooldown or shutdown: Location-Modes- Modes-Safe Shutdown Area 1, 2 3,4, 5-17 North RHR Unit-1 RHR Equipment. RHR Shut Down Cooling (SDC)-No entry required RHR Pump Discharge Isolation Valves Ell-FO18 A&C Inventory Control Equipment-No entry required Reactivity Control.-No entry required-17 North RHR Unit-2 RHR Equipment. RHR Shut Down Cooling (SDC)-No entry required RHR Pump Discharge Isolation Valves El 1-FO18 A&C Inventory Control Equipment. -No entry required Reactivity Control.-No entry required-17 South RHR Unit-1 RHR Equipment. RHR Shut Down Cooling (SDC)-No entry required RHR Pump Discharge Isolation Valves Ell-FO18 B&D Inventory Control Equipment. -No entry required Reactivity Control.-No entry required-17 South RHR Unit-2 RHR Equipment. RHR Shut Down Cooling (SDC)-No entry required RHR Pump Discharge Isolation Valves El -F018 B&D Inventory Control Equipment. -No entry required Reactivity Control.-No entry required-17 North Core Spray Core Spray Equipment Inventory Control-No entry required -No entry required-17 South Core Spray Core Spray Equipment Inventory Control.-No entry required -No entry required Service Water Building 20' Heat Sink equipment. Heat Sink equipment. -No entry required No entry required OPEP-02.2.1 Rev. 6 Page 307 of 310 ATTACHMENT 3 Page 3 of 4 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases Control Building HVAC Room Habitability. Habitability.(Turbine Building 70') -No entry required -No entry required Reactor Building HVAC Habitability. Habitability.(Reactor Building 80' West) -No entry required -No entry required Electrical Power, Local control Electrical Power, Local control Emergency Diesel Generators parameters at EDG panel. parameters at EDG panel.(EDG Building 20') No entry required -No entry required Electrical Power. Electrical Power.4-Day Tank Rooms -No entry required -No entry required EDG Building HVAC (EDG Habitability. Habitability. Building 70')- No entry required -No entry required Electrical Power. Electrical Power.1 &2) No entry required No entry required 4160 VAC (EDG building 70') No entry required No entry required 480 VAC (EDG Building 20' )N No entry required No entry required& S ends 120 VAC Vital (Cable Spread U- No entry required No entry required 1 & U-2)Train A & B DC (Battery Rooms No entry required No entry required U-1 & U-2)Reactor Building 20' East & No entry required RHR SDC.West MCC Areas -1 (2) El1 -F009 & F008 valve breakers (RHR SDC Suction isolation valves)-1 (2) Eli -F006 A-D valve breakers, (RHR pump suction valves).RHR SDC Suction Fill &Vent valves (Manual Valves)-RHR suction pipe flush valve breakers (El 1-FO11A & B. E11-V33. Ell-Reactor Building 20' Pipe -No entry required RHR SDC Tunnel -RHR SDC Suction Fill &Vent valves (Manual Valves)OPEP-02.2.1 Rev. 6 Page 308 of 310 ATTACHMENT 3 Page 4 of 4 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases Table R-2 & H-2 Results Table R-2 & H-2 Safe Operation & Shutdown Areas Room/Area Mode Applicability Reactor Building -17' North RHR Unit-1 & 2 3, 4, 5 Reactor Building -17' South RHR Unit-i & 2 3, 4, 5 Reactor Building 20' East & West MCC Areas Unit-1 & 2 3, 4, 5 Reactor Building 20' Pipe Tunnel Unit-1 & 2 3, 4, 5 Plant Operating Procedures Reviewed 1. Unit 1 & 2 RHR OP-17" Shutdown Cooling" Low Pressure Coolant Injection 2.3.4.5.6.7.Unit-1 & 2 UAT Backfeed OP-50 EDG Operation OP-50.1, OP-39 Unit-1 & 2 Service Water OP-43 Unit-i & 2 Core Spray Control Building Ventilation System 20P-37 Defense In Depth AP-22 IOPEP-02.2.1 I Rev. 6 Page 309 of 310 REVISION

SUMMARY

Revision 6 of OPEP-02.2.1 consists of the following changes:[xxxx]IOPEP-02.2.1 I Rev. 6 1 Page 310 of 310 1 Page 163: [1] Deleted Author EAL #4 If a FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected Area] PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. BSEP 15-0010 Enclosure 3 Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 License Amendment Request to Adopt Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors" BSEP Technical Bases Document, OPEP-02.2.1,"Emergency Action Level Technical Bases" (Clean Version) ENERGY. BRUNSWICK NUCLEAR PLANT Reference Use PLANT OPERATING MANUAL VOLUME XIII PLANT EMERGENCY PROCEDURE OPEP-02.2.1 EMERGENCY ACTION LEVEL TECHNICAL BASES REVISION 6 (Draft D4 8/16/14)IOPEP-02.2.1 I Rev. 6 Page 1 of 303 TABLE OF CONTENTS SECTION PAGE 1 .0 P U R P O S E ................................................................................................................................... 3 2.0 DISCUSSION ............................................................................................................................... 3 2 .1 B a c kg ro u n d ............................................................................................................................... 3 2.2 Fission Product Barriers ......................................... ............ 4 2.3 Fission Product Barrier Classification Criteria ...................................................................... 4 2.4 EAL Organization ...................................................................................................................... 5 2.5 Technical Bases Inform ation ............................................................................................... 7 2.6 Operating Mode Applicability ............................................................................................... 8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ............................................... 9 3.1 General Considerations ........................................................................................................ 9 3.2 Classification Methodology ................................................................................................. 10

4.0 REFERENCES

........................................................................................................................... 14 4.1 Developmental ........................................................................................................................ 14 4 .2 Im p le m e n tin g ....................................................... ...................................................................... 5.0 DEINITIONS, ACRONYM S & ABBREVIATIONS .................................................................. 15 6.0 BNP TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE ......................................................... 22 7.0 ATTACHM ENTS ........................................................................................................................ 26 1 Emergency Action Level Technical Bases ............................................................... 27 Category R Abnormal Rad Release / Rad Effluent .......................................... 27 Cate-gory C Cold Shutdown / Refueling System Malfunction ........................... 74 Category H Hazards ........................................................................................... 127 Category S System M alfunction ......................................................................... 178 Category E ISFSI ............................................................................................... 227 Category F Fission Product Barrier Degradation ................................................ 231 2 Fission Product Barrier Loss / Potential Loss Matrix and Bases ......................................................................................................... 237 3 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases ...................................... 299 I OPEP-02.2.1 Rev. 6 1 Page 2 of 303 1

1.0 PURPOSE

This document provides an explanation and rationale for each Emergency Action Level (EAL)included in the EAL Upgrade Project for Brunswick Nuclear Plant (BNP). It should be used to facilitate review of the BNP EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of OPEP-02.1 Initial Emergency Actions, may use this document as a technical reference in support of EAL interpretation. This information may assist the Site Emergency Coordinator in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to offsite officials. The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.

2.0 DISCUSSION

2.1 Background

EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the BNP Emergency Plan.In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance.NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included: " Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.

Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSls)." Simplifying the fission product barrier EAL threshold for a Site Area Emergency.

Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs).Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ADAMS Accession Number ML1 10240324) (ref.4.1.1), BNP conducted an EAL implementation upgrade project that produced the EALs discussed herein I 0PEP-02.2.1 Rev. 6 Page 3 of 303

2.2 Fission

Product Barriers 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. Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. "Loss" and"Potential Loss" signify the relative damage and threat of damage to the barrier. A "Loss" threshold means the barrier no longer assures containment of radioactive materials. A"Potential Loss" threshold implies an increased probability of barrier loss and decreased certainty of maintaining the barrier.The primary fission product barriers are: A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.B. Reactor Coolant System (RCS): 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.C. Primary Containment (PC): 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. Primary Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.

2.3 Fission

Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss: Alert: Any loss or any potential loss of either Fuel Clad or RCS barrier Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of the third barrier OPEP-02.2.1 Rev. 6 Page 4 of 303] 2.4 EAL Organization The BNP EAL scheme includes the following features: " Division of the EAL set into three broad groups: o EALs applicable under all plant operating modes -This group would be reviewed by the EAL-user any time emergency classification is considered. o EALs applicable only under hot operating modes -This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup, or Power Operation mode.o EALs applicable only under cold operating modes -This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode.The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency." Within each group, assignment of EALs to categories and subcategories: Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The BNP EAL categories are aligned to and represent the NEI 99-01"Recognition Categories." Subcategories are used in the BNP scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The BNP EAL categories and subcategories are listed below.OPEP-02.2.1 Rev. 6 Page 5 of 303] EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory Any Operating Mode: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent 2 -Irradiated Fuel Event 3- Area Radiation Levels H -Hazards and Other Conditions 1 -Security Affecting Plant Safety 2 -Seismic Event 3 -Natural or Technological Hazard 4 -Fire 5 -Hazardous Gas 6 -Control Room Evacuation 7 -SEC Judgment E -Independent Spent Fuel Storage 1 -Confinement Boundary Installation (ISFSI)Hot Conditions: S -System Malfunction 1 -Loss of Emergency AC Power 2 -Loss of Vital DC Power 3 -Loss of Control Room Indications 4 -RCS Activity 5 -RCS Leakage 6 -RPS Failure 7 -Loss of Communications 8 -Hazardous Event Affecting Safety Systems F -Fission Product Barrier Degradation None Cold Conditions: C -Cold Shutdown / Refueling System 1 -RPV Level Malfunction 2 -Loss of Emergency AC Power 3 -RCS Temperature 4 -Loss of Vital DC Power 5 -Loss of Communications 6 -Hazardous Event Affecting Safety Systems The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL Technical Bases Document in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachments 1 & 2 of this document for such information. OPEP-02.2.1 Rev. 6 Page 6 of 303]

2.5 Technical

Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (R, C, H, S, F and E) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided: Cate-gory Letter & Title Subcategory Number & Title Initiating Condition (IC)Site-specific description of the generic IC given in NEI 99-01 Rev. 6.EAL Identifier (enclosed in rectangle) Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:

1. First character (letter):

Corresponds to the EAL category as described above (R, C, H, S, F or E)2. Second character (letter): The emergency classification (G, S, A or U)G = General Emergency S = Site Area Emergency A = Alert U = Unusual Event 3. Third character (number): Subcategory number within the given category.Subcategories are sequentially numbered beginning with the number one (1). If a category does not have a subcategory, this character is assigned the number one (1).4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1).Classification (enclosed in rectangle): Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G)EAL (enclosed in rectangle) Exact wording of the EAL as it appears in the EAL Classification Matrix Mode Applicability OPEP-02.2.1 Rev. 6 Page 7 of 303 One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown, 4 -Cold Shutdown, 5 -Refueling, D -Defueled, or All. (See Section 2.6 for operating mode definitions) Definitions: If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1.Basis: A Plant-Specific basis section that provides BNP-relevant information concerning the EAL.This is followed by a Generic basis section that provides a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.BNP Basis Reference(s): Site-specific source documentation from which the EAL is derived 2.6 Operating Mode Applicability (ref. 4.1.7)1 Power Operations Reactor is critical and the mode switch is in RUN 2 Startup The mode switch is in STARTUP/HOT STANDBY 3 Hot Shutdown The mode switch is in SHUTDOWN, all reactor vessel head closure bolts are fully tensioned, and reactor coolant temperature is >212*F 4 Cold Shutdown The mode switch is in SHUTDOWN, all reactor vessel head closure bolts are fully tensioned, and reactor coolant temperature is < 212*F 5 Refuel The mode switch is in REFUEL or SHUTDOWN and one or more reactor vessel head closure bolts are less than fully tensioned D Defueled RPV contains no irradiated fuel The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.OPEP-02.2.1 Rev. 6 Page 8 of 303]

3.0 GUIDANCE

ON MAKING EMERGENCY CLASSIFICATIONS

3.1 General

Considerations When making an emergency classification, the Site Emergency Coordinator 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 based on loss or potential loss of Fission Product Barrier Thresholds.

3.1.1 Classification

Timeliness 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" (ref. 4.1.12).3.1.2 Valid Indications All emergency classification assessments shall 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, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

3.1.3 Imminent

Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Site Emergency Coordinator 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.3.1.4 Planned vs. Unplanned Events 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 OPEP-02.2.1 Rev. 6 1 Page 9 of 303 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 (ref.4.1.4).3.1.5 Classification Based on Analysis 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.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift). 3.1.6 Site Emergency Coordinator Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Site Emergency Coordinator 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 Site Emergency Coordinator 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 in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.3.2 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 must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process "clock" starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process "clock" started.When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock." For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 4.1.14).3.2.1 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: OPEP-02.2.1 Rev. 6 Page 10 of 303]

If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two different units, a Site Area Emergency should be declared.There is no "additive" effect from multiple EALs meeting the same ECL. For example:* If two Alert EALs are met, whether at one unit or at two different units, an Alert should be declared.Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.2).3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable.

If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher.3.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Site Emergency Coordinator 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 Site Emergency Coordinator, 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.3.2.4 Emergency Classification Level Upgrading and Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s)and EAL(s). The ECL may also simply be terminated. As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.2).3.2.5 Classification of Short-Lived Events OPEP-02.2.1 I Rev. 6 Page 11 of303 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 the reactor followed by a successful manual scram.3.2.6 Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions. EAL momentarily met during expected plant response -In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures. EAL momentarily met but the condition is corrected prior to an emergency declaration -If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example: An ATWS occurs and the high pressure ECCS systems fail to automatically start. RPV level rapidly decreases and the plant enters an inadequate core cooling condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be based on the ATWS only.It is important to stress that the 15-minute emergency classification assessment period (process clock) 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 when an operator is able to take a successful corrective action prior to the Site Emergency Coordinator 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. 3.2.7 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 OPEP-02.2.1 Rev. 6 1 Page 12 of 303] condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 (ref. 4.1.4) 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.

3.2.8 Retraction

of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3).OPEP-02.2.1 Rev. 6 Page 13 of 303]

4.0 REFERENCES

4.1 Developmental

4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML1 10240324 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007.4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 § CFR 50.73 License Event Report System 4.1.6 BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan 4.1.7 BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations 4.1.8 Technical Specifications Table 1.1-1 Modes 4.1.9 Technical Specifications Section 3.6 Containment Systems 4.1.10 PRO-NGGC-0201 NGG Procedure Writers Guide 4.1.11 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.12 NGGM-PM-0028 Transnuclear NUHOMS Dry Fuel Storage Program Manual 4.2 Implementing 4.2.1 OPEP-02.1 Initial Emergency Actions 4.2.2 NEI 99-01 Rev. 6 to BNP EAL Comparison Matrix 4.2.3 BNP EAL Matrix I OPEP-02.2.1 Rev. 6 Page 14 of 303 1 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS

5.1 Definitions

(ref. 4.1.1 except as noted)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.Can/Cannot Be Maintained Above/Below The value of an identified parameter is/is not able to be held within the specified limit. The determination requires an evaluation of system performance and availability in relation to parameter values and trends. An instruction prescribing action when a parameter cannot be maintained above or below a specified limit neither requires nor prohibits anticipatory action-depending upon plant conditions, the action may be taken as soon as it is determined that the limit will ultimately be exceeded, or delayed until the limit is actually reached. Once the parameter does exceed the limit, however, the action must be performed; it may not be delayed while attempts are made to restore the parameter to within the desired control band.Can/Cannot Be Restored Above/Below The value of an identified parameter is/is not able to be brought within the specified limit. The determination requires an evaluation of system performance and availability in relation to parameter values and trends. An instruction prescribing action when a value cannot be restored and maintained above or below a specified limit does not require immediate action simply because the current values is outside the range, but does not permit extended operation beyond the limit; the action must be taken as soon as it is apparent that the specified range cannot be attained.Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the BNP ISFSI, Confinement Boundary is defined as the Dry Shielded Canister (DSC) (Ref. 4.1.12).Containment Closure The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to BNP, Containment Closure is established when either Primary Containment is Operable per Section 3.6.1.1 of Technical Specifications or Secondary Containment is considered functional per the requirements of OAP-022, BNP Outage Risk Management. EPA PAGs Environment Protection Agency Protective Action Guidelines. The EPA PAGs are expressed in terms of dose commitment: 1 Rem TEDE or 5 Rem CDE Thyroid. Actual or projected offsite OPEP-02.2.1 Rev. 6 Page 15 of 303] exposures in excess of the EPA PAGs requires BNP to recommend protective actions for the general public to offsite planning agencies.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 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.Flooding A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.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 BNP 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 BNP. 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.Impede(d)OPEP-02.2.1 Rev. 6 Page 16 of 303] Personnel access to a room or area is hindered to an extent that 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). Intrusion The act of entering without authorization. Discovery of a bomb in a specified area is indication of intrusion into that area by a hostile force.Maintain Take appropriate action to hold the value of an identified parameter within specified limits.Normal Levels As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value.Owner Controlled Area Area depicted as the property boundary in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan (ref. 4.1.6).Projectile An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.Protected Area The double-fenced security area with intrusion detection devices immediately surrounding the plant structures. The Protected Area is depicted in BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations (ref. 4.1.7).RCS Intact The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).Refueling Pathway The reactor refueling cavity, spent fuel pool and fuel transfer canal (cattle chute) comprise the refueling pathway.Restore Take the appropriate action required to return the value of an identified parameter to the applicable limits Safety System OPEP-02.2.1 Rev. 6 Page 17 of 303] 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 (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. 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.Site Boundary Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan (ref. 4.1.6).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.Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. 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. I OPEP-02.2.1 Rev. 6 1 Page 18 of 303 5.2 Abbreviations/Acronyms OF ....................................................................................................... Degrees Fahrenheit 0 ........................................................................................................................... D e g re e s AC ....................................................................................................... Alternating Current AO P ................................................................................. Abnorm al O perating Procedure APRM .................................................................................. Average Power Range M eter ATW S ...................................................................... Anticipated Transient W ithout Scram BNP ............................................................................................ Brunswick Nuclear Plant BW R ............................................................................................... Boiling W ater Reactor BW RO G .................................................................. Boiling W ater Reactor Owners G roup CDE ....................................................................................... Com m itted Dose Equivalent CFR ..................................................................................... Code of Federal Regulations CS ................................................................................................................... Core Spray DBA ............................................................................................... Design Basis Accident DC ............................................................................................................... Direct Current EAL ............................................................................................. Em ergency Action Level ECCS ............................................................................ Em ergency Core Cooling System ECL .................................................................................. Em ergency Classification Level EO F .................................................................................. Em ergency Operations Facility EO P ............................................................................... Em ergency Operating Procedure EPA .............................................................................. Environm ental Protection Agency EPG ............................................................................... Em ergency Procedure G uideline EPIP ................................................................ Em ergency Plan Im plem enting Procedure ESF ......................................................................................... Engineered Safety Feature FAA .................................................................................. Federal Aviation Adm inistration FBI ................................................................................... Federal Bureau of Investigation FEM A ............................................................... Federal Em ergency M anagem ent Agency FSAR .................................................................................... Final Safety Analysis Report G E ..................................................................................................... General Em ergency HCTL ............................................................................ Heat Capacity Tem perature Lim it HPCI ................................................................................ High Pressure Coolant Injection IC .......................................................................................................... Initiating Condition IPEEE ................. Individual Plant Examination of External Events (Generic Letter 88-20)ISFSI ............................................................ Independent Spent Fuel Storage Installation OPEP-02.2.1 Rev. 6 Page 19 of 303 Keff ......................................................................... Effective Neutron M ultiplication Factor LCO .................................................................................. Lim iting Condition of Operation LER ................................................................................................ Licensee Event Report LO CA ......................................................................................... Loss of Coolant Accident LPSI .................................................................................... Low Pressure Safety Injection LW R ................................................................................................... Light W ater Reactor M PC ................................... M axim um Perm issible Concentration/M ulti-Purpose Canister M PH ........................................................................................................... M iles Per Hour M SIV ....................................................................................... M ain Steam Isolation Valve M SL ........................................................................................................ M ain Steam Line m R, m Rem , m rem , m REM .............................................. m illi-Roentgen Equivalent M an M W .................................................................................................................... M egawatt NEI .............................................................................................. Nuclear Energy Institute NESP ................................................................... National Environm ental Studies Project NPP .................................................................................................. Nuclear Power Plant NRC ................................................................................ Nuclear Regulatory Com m ission NSSS ................................................................................ Nuclear Steam Supply System NO RAD ................................................... North Am erican Aerospace Defense Com m and (NO)UE ................................................................................ Notification of Unusual Event O BE ...................................................................................... O perating Basis Earthquake OCA ............................................................................................... Ow ner Controlled Area O DCM/O DAM .......................................... Offsite Dose Calculation (Assessm ent) M anual O RO ................................................................................. Offsite Response O rganization PA .............................................................................................................. Protected Area PRA/PSA ..................... Probabilistic Risk Assessment / Probabilistic Safety Assessment PW R ....................................................................................... Pressurized W ater Reactor PSIG ................................................................................ Pounds per Square Inch Gauge R ........................................................................................................................ Roentgen RB ........................................................................................................... Reactor Building RCIC ................................................................................. Reactor Core Isolation Cooling RCS ............................................................................................ Reactor Coolant System Rem , rem , REM ....................................................................... Roentgen Equivalent M an RETS ......................................................... Radiological Effl uent Technical Specifications RPS ........................................................................................ Reactor Protection System OPEP-02.2.1 Rev. 6 Page 20 of 303 RPV ........................................................................................... Reactor Pressure Vessel RW CU .......................................................................................... Reactor W ater Cleanup SAR ............................................................................................... Safety Analysis Report SBGTS ......................................................................... Stand-By G as Treatm ent System SBO ......................................................................................................... Station Blackout SCBA ....................................................................... Self-Contained Breathing Apparatus SEC ...................................................................................... Site Em ergency Coordinator SPDS ........................................................................... Safety Param eter Display System SRO ............................................................................................ Senior Reactor O perator TEDE ............................................................................... Total Effective Dose Equivalent TAF ....................................................................................................... Top of Active Fuel TSC .......................................................................................... Technical Support Center OPEP-02.2.1 Rev. 6 Page 21 of 303] 6.0 BNP-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a BNP EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the BNP EALs based on the NEI guidance can be found in the EAL Comparison Matrix.BNP NEI 99-01 Rev. 6 EAL IC Example EAL RU1.1 AU1 1,2 RU1.2 AU1 3 RU2.1 AU2 1 RA1.1 AA1 1 RA1.2 AA1 2 RA1.3 AA1 3 RA1.4 AA1 4 RA2.1 AA2 1 RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1 RA3.2 AA3 2 RS1.1 AS1 1 RS1.2 AS1 2 RS1.3 AS1 3 RS2.1 AS2 1 RG1.1 AG1 1 RG1.2 AG1 2 RG1.3 AG1 3 RG2.1 AG2 1 OPEP-02.2.1 Rev. 6 Page 22 of 303] BNP NEI 99-01 Rev. 6 EAL IC Example EAL CU1.1 CUI 1 CU1.2 CU1 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1,2, 3 CAl.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1,2 CA6.1 CA6 1 CSl.1 CSl 1 CSl.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 1 CG1.2 CG1 2 FA1.1 FA1 1 FS1.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1 HU1.2 HUI 2 HU1.3 HUI 3 HU2.1 HU2 1 HU3.1 HU3 1 OPEP-02.2.1 Rev. 6 Page 23 of 303] BNP NEI 99-01 Rev. 6 EAL IC Example EAL HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 HU3.5 HU3 5 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HAl 1 HA1.2 HAl 2 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA7 1 HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 HG1.1 HG1 1 HG7.1 HG7 1 SU1.1 SUl 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1,2,3 SU6.1 SU5 1 OPEP-02.2.1 Rev. 6 Page 24 of 303] BNP NEI 99-01 Rev. 6 EAL IC Example EAL SU6.2 SU5 2 SU7.1 SU6 1,2,3 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA8.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SG8 1 EU1.1 E-HU1 1 OPEP-02.2.1 Rev. 6 Page 25 of 303]

7.0 ATTACHMENTS

7.1 7.2 Attachment 1, Emergency Action Level Technical Bases Attachment 2, Fission Product Barrier Matrix and Basis I OPEP-02.2.1 Rev. 6 1 Page 26 of 303 1 ATTACHMENT 1 Page 1 of 210 EAL Bases Category R -Abnormal Rad Release I Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification. At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.Events of this category pertain to the following subcategories:

1. Radiolo-iical Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.

3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.

I OPEP-02.2.1 Rev. 6 Page 27 of 303 1 ATTACHMENT 1 Page 2 of 210 EAL Bases Category: Subcategory: Initiating Condition: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Release of gaseous or liquid radioactivity > 2 times the ODCM limits for 60 minutes or longer EAL: RUI.1 Unusual Event Reading on any Table R-1 effluent radiation monitor > column "UE" for 2t 60 min.(Notes 1,2, 3)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE [ Alert UE Main Stack Rad D12-RM-23S 2.13E+09 pCi/sec 2.13E+08 pCi/sec 2.13E+07 pCi/sec 1.80E+06 pCi/sec Reactor Bldg Vent Noble cAc-AQH-1264-3 ..... 6.14E+04 cpm SGas'U Turbine Bldg Vent D12-RM-23 1.07E+08 pCi/sec 1.07E+07 pCi/sec 1.07E+06 pCi/sec 1.13E+04 pCi/sec Service Water Effluent D12-RM-K605

2 x hi alarm Rad E-I Radwaste Effluent Rad D12-RM-K604

....... 2 x hi-hi alarm Mode Applicability: All OPEP-02.2.1 Rev. 6 Page 28 of 303] ATTACHMENT 1 Page 3 of 210 EAL Bases Definition(s): None BNP Basis: Gaseous Releases The column "UE" gaseous release values in Table R-1 represent two times the appropriate ODCM release rate limits associated with the specified monitors (ref. 2, 3).Instrumentation that may be used to assess this EAL is listed below (ref. 1): " Main Stack Monitoring System Noble Gas Activity Monitor 2-D12-RM-23S (1/2-D12-RR-4599-4)* Reactor Building Ventilation Monitoring System Noble Gas Activity Monitor 1(2)-CAC-AQH-1264-3

Turbine Building Ventilation Monitoring System Noble Gas Activity Monitor 1(2)-D12-RM-23 (1/2-DL12-RR-4548-4)

Liquid Releases Instrumentation that may be used to assess this EAL is listed below: " Liquid Radwaste Radioactivity Monitor 2-D12-RM-K604 (batch release)* Main Service Water Effluent Radioactivity Monitor 1(2)-D12-RM-K605 (continuous release)The Liquid Radwaste Radioactivity Monitor Hi-Hi alarm automatically closes Radwaste Liquid Effluent Discharge Valves D12-V27A and 27B. The Hi-Hi alarm setpoint is set in accordance with the ODCM and includes a conservative reduction factor of 20 to the ODCM release rate limit (ref. 1, 2).The Main Service Water Effluent Radioactivity Monitor High alarm setpoint is set in accordance with the ODCM and ensures continuous liquid releases do not exceed ODCM Section 7.3.3 limits.OPEP-02.2.1 Rev. 6 Page 29 of 303] ATTACHMENT 1 Page 4 of 210 EAL Bases NEI 99-01 Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.Nuclear 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.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.This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.Escalation of the emergency classification level would be via IC RAI.I OPEP-02.2.1 Rev. 6 1 Page 30 of 303] ATTACHMENT 1 Page 5 of 210 EAL Bases BNP Basis Reference(s):

1. BNP ODCM Appendix E Radioactive Liquid and Gaseous Effluent Monitoring Instrumentation Numbers 2. BNP Offsite Dose Calculation Manual 3. EP-EALCALC-BNP-0801 Radiological Gaseous Effluent Values (EALs RG1, RS1, RA1 and RU 1)4. NEI 99-01 AU1 OPEP-02.2.1 Rev. 6 Page 31 of 303]

ATTACHMENT 1 Page 6 of 210 EAL Bases Category: Subcategory: Initiating Condition: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.EAL: RU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate> 2 x ODCM limits for > 60 min. (Notes 1, 2)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Mode Applicability: All Definition(s): None BNP Basis: None NEI 99-01 Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.Nuclear 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.OPEP-02.2.1 Rev. 6 Page 32 of 303] ATTACHMENT 1 Page 7 of 210 EAL Bases 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. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).Escalation of the emergency classification level would be via IC RAI.BNP Basis Reference(s):

1. BNP ODCM Appendix E Radioactive Liquid and Gaseous Effluent Monitoring Instrumentation Numbers 2. BNP Offsite Dose Calculation Manual 3. NEI 99-01 AU1 OPEP-02.2.1 Rev. 6 Page 33 of 303]

ATTACHMENT 1 Page 8 of 210 EAL Bases Category: Subcategory: Initiating Condition: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: RA1.1 Alert In the absence of real-time dose assessment, reading on any Table R-1 effluent radiation monitor > column "ALERT" for > 15 min. (Notes 1, 2, 3, 4)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4 The pre-calculated effluent monitor values presented in EALs RAI.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Main Stack Rad D12-RM-23S 2.13E+09 pCi/sec 2.13E+08 pCi/sec 2.13E+07 pCi/sec 1.80E+06 pCi/sec Reactor Bldg Vent Noble 0GsCAC-AQH-1264-3

-- 6.14E+04 cpm oGas (U Turbine Bldg Vent D12-RM-23 1.07E+08 pCi/sec 1.07E+07 pCi/sec 1.07E+06 pCi/sec 1.13E+04 pCi/sec Service Water Effluent D12-RM-K605 2 x hi alarm Rad Radwaste Effluent Rad D12-RM-K604

.....----

2 x hi-hi alarm OPEP-02.2.1 Rev. 6 Page 34 of 303] ATTACHMENT 1 Page 9 of 210 EAL Bases Mode Applicability: All Definition(s): None BNP Basis: This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either (ref. 2, 3):* 10 mRem TEDE* 50 mRem CDE Thyroid The column "ALERT" gaseous effluent release values in Table R-1 correspond to calculated doses of 1% (10% of the SAE thresholds) of the EPA Protective Action Guidelines (TEDE or CDE Thyroid).Instrumentation that may be used to assess this EAL is listed below (ref 1): " Main Stack Monitoring System Noble Gas Activity Monitor 2-D12-RM-23S (1/2-D12-RR-4599-4)* Turbine Building Ventilation Monitoring System Noble Gas Activity Monitor 1(2)-D12-RM-23 (1/2-DL12-RR-4548-4) NEI 99-01 Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.OPEP-02.2.1 Rev. 6 1 Page 35 of 303 ATTACHMENT 1 Page 10 of 210 EAL Bases Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Escalation of the emergency classification level would be via IC RS1.BNP Basis Reference(s):

1. BNP ODCM Appendix E Radioactive Liquid and Gaseous Effluent Monitoring Instrumentation Numbers 2. BNP Offsite Dose Calculation Manual 3. EP-EALCALC-BNP-0801 Radiological Gaseous Effluent Values (EALs RG1, RS1, RA1 and RU 1)4. NEI 99-01 AA1 IOPEP-02.2.1 I Rev. 6 1 Page 36 of 303 ATTACHMENT 1 Page 11 of 210 EAL Bases Category: Subcategory:

Initiating Condition: R -Abnormal Rad Levels I Rad Effluent 1 -Radiological Effluent Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: RA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 3, 4)Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All Definition(s): SITE BOUNDARY-Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan BNP Basis: Dose assessments are performed by computer-based methods (ref. 1, 2, 3)NEI 99-01 Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).I OPEP-02.2.1 Rev. 6 Page 37 of 303] ATTACHMENT 1 Page 12 of 210 EAL Bases Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Escalation of the emergency classification level would be via IC RS1.BNP Basis Reference(s):

1. EMG-NGGC-0002 Off-site Dose Assessment

2. OPEP-03.4.7 Automation of Off-Site Dose Projections

3. OE&RC-03.4.8, Offsite Dose Projections for Monitored Releases 3. NEI 99-01 AA1 OPEP-02.2.1 Rev. 6 Page 38 of 303]

ATTACHMENT 1 Page 13 of 210 EAL Bases Category: Subcategory: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: RA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Mode Applicability: All Definition(s): SITE BOUNDARY-Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan BNP Basis: Dose assessments based on liquid releases are performed per Offsite Dose Calculation Manual (ref. 1).OPEP-02.2.1 Rev. 6 Page 39 of 303] ATTACHMENT 1 Page 14 of 210 EAL Bases NEI 99-01 Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Escalation of the emergency classification level would be via IC RS1.BNP Basis Reference(s):

1. BNP Offsite Dose Calculation Manual 2. NEI 99-01 AA1 OPEP-02.2.1 Rev. 6 Page 40 of 303]

ATTACHMENT 1 Page 15 of 210 EAL Bases Category: R -Abnormal Rad Levels / Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: RAI.4 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: " Closed window dose rates > 10 mR/hr expected to continue for 2- 60 min." Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation.(Notes 1,2)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Mode Applicability: All Definition(s): SITE BOUNDARY-Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan BNP Basis: OPEP-02.6.6, Environmental Monitoring Team Leader and OPEP-03.5.5 Environmental Monitoring and Plume Tracking provide guidance for emergency or post-accident radiological environmental monitoring (ref. 1,2).OPEP-02.2.1 Rev. 6 Page 41 of 303] ATTACHMENT 1 Page 16 of 210 EAL Bases NEI 99-01 Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Escalation of the emergency classification level would be via IC RS1.BNP Basis Reference(s):

1. OPEP-02.6.6, Environmental Monitoring Team Leader 2. OPEP-03.5.5 Environmental Monitoring and Plume Tracking 3. NEI 99-01 AA1 I OPEP-02.2.1 Rev. 6 1 Page 42 of 303 1 ATTACHMENT 1 Page 17 of 210 EAL Bases Category: Subcategory:

R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL: RSI.1 Site Area Emergency In the absence of real-time dose assessment, reading on any Table R-1 effluent radiation monitor > column "SAE" for > 15 min.(Notes 1,2, 3, 4)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Main Stack Rad D12-RM-23S 2.13E+09 pCi/sec 2.13E+08 pCi/sec 2.13E+07 pCi/sec 1.80E+06 pCi/sec Reactor Bldg Vent Noble SGas CAC-AQH-1264-3

.----

-- 6.14E+04 cpm Turbine Bldg Vent D12-RM-23 1.07E+08 pCi/sec 1.07E+07 pCi/sec 1.07E+06 pCi/sec 1.13E+04 pCi/sec Service Water Effluent D12-RM-K605

2 x hi alarm Rad Radwaste Effluent Rad D12-RM-K604 2 x hi-hi alarm Mode Applicability:

All OPEP-02.2.1 Rev. 6 Page 43 of 303 ATTACHMENT 1 Page 18 of 210 EAL Bases Definition(s): None BNP Basis: This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either (ref. 2): 0 100mRemTEDE

500 mRem CDE Thyroid The column "SAE" gaseous effluent release values in Table R-1 correspond to calculated doses of 10% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid).Instrumentation that may be used to assess this EAL is listed below (ref 1):* Main Stack Monitoring System Noble Gas Activity Monitor 2-D12-RM-23S (1/2-D12-RR-4599-4)" Turbine Building Ventilation Monitoring System Noble Gas Activity Monitor 1(2)-D12-RM-23 (1/2-DL12-RR-4548-4)

NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that 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 RG1.OPEP-02.2.1 Rev. 6 Page 44 of 303] ATTACHMENT 1 Page 19 of 210 EAL Bases BNP Basis Reference(s):

1. BNP ODCM Appendix E Radioactive Liquid and Gaseous Effluent Monitoring Instrumentation Numbers 2. EP-EALCALC-BNP-0801 Radiological Gaseous Effluent Values (EALs RG1, RS1, RA1 and RU1)3. NEI 99-01 AS1 OPEP-02.2.1 Rev. 6 Page 45 of 303]

ATTACHMENT 1 Page 20 of 210 EAL Bases Category: Subcategory: Initiating Condition: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL: RS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 3, 4)Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All Definition(s): SITE BOUNDARY -Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan BNP Basis: Dose assessments are performed by computer-based methods (ref. 1, 2, 3)NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.I OPEP-02.2.1 Rev. 6 Page 46 of 303 ATTACHMENT 1 Page 21 of 210 EAL Bases 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 RG1.BNP Basis Reference(s):

1. EMG-NGGC-0002 Off-site Dose Assessment

2. OPEP-03.4.7 Automation of Off-Site Dose Projections

3. OE&RC-03.4.8, Offsite Dose Projections for Monitored Releases 4. NEI 99-01 AS1 IOPEP-02.2." I Rev. 6 Page 47 of 303 ATTACHMENT 1 Page 22 of 210 EAL Bases Category: Subcategory:

Initiating Condition: R -Abnormal Rad Levels I Rad Effluent 1 -Radiological Effluent Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL: RS1.3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:* Closed window dose rates > 100 mR/hr expected to continue for > 60 min." Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation.(Notes 1,2)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Mode Applicability: All Definition(s): SITE BOUNDARY -Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan BNP Basis: OPEP-02.6.6, Environmental Monitoring Team Leader and OPEP-03.5.5 Environmental Monitoring and Plume Tracking provide guidance for emergency or post-accident radiological environmental monitoring (ref. 1, 2).OPEP-02.2."I Rev. 6 Page 48 of 303] ATTACHMENT 1 Page 23 of 210 EAL Bases NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that 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.Escalation of the emergency classification level would be via IC RG1.BNP Basis Reference(s):

1. 0PEP-02.6.6, Environmental Monitoring Team Leader 2. 0PEP-03.5.5 Environmental Monitoring and Plume Tracking 3. NEI 99-01 AS1 OPEP-02.2.1 Rev. 6 Page 49 of 303]

ATTACHMENT 1 Page 24 of 210 EAL Bases Category: Subcategory: Initiating Condition: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL: RG1.1 General Emergency In the absence of real-time dose assessment, reading on any Table R-1 effluent radiation monitor > column "GE" for -15 min.(Notes 1,2, 3, 4)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Main Stack Rad D12-RM-23S 2.13E+09 pCi/sec 2.13E+08 pCi/sec 2.13E+07 pCi/sec 1.80E+06 pCi/sec Reactor Bldg Vent Noble cAc-AQH-1264-3 6.14E+04 cpm Gas Turbine Bldg Vent D12-RM-23 1.07E+08 pCi/sec 1.07E+07 pCi/sec 1.07E+06 pCi/sec 1.13E+04 pCi/sec Service Water Effluent D12-RM-K605 2 x hi alarm Rad Radwaste Effluent Rad D12-RM-K604 .... 2 x hi-hi alarm 0PEP-02.2.1 Rev. 6 Page 50 of 303] ATTACHMENT 1 Page 25 of 210 EAL Bases Mode Applicability: All Definition(s): None BNP Basis: This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either (ref. 2): 0 1000mRemTEDE

5000 mRem CDE Thyroid The column "SAE" gaseous effluent release values in Table R-1 correspond to calculated doses of 100% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid).Instrumentation that may be used to assess this EAL is listed below (ref 1): " Main Stack Monitoring System Noble Gas Activity Monitor 2-D12-RM-23S (1/2-D12-RR-4599-4)* Turbine Building Ventilation Monitoring System Noble Gas Activity Monitor 1(2)-D12-RM-23 (1/2-DL12-RR-4548-4)

NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.OPEP-02.2.1 Rev. 6 Page 51 of 303] ATTACHMENT 1 Page 26 of 210 EAL Bases 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.BNP Basis Reference(s):

1. BNP ODCM Appendix E Radioactive Liquid and Gaseous Effluent Monitoring Instrumentation Numbers 2. EP-EALCALC-BNP-0801 Radiological Gaseous Effluent Values (EALs RG1, RS1, RA1 and RU1)3. NEI 99-01 AGI I OPEP-02.2.1 Rev. 6 Page 52 of 303]

ATTACHMENT 1 Page 27 of 210 EAL Bases Category: Subcategory: R -Abnormal Rad Levels I Rad Effluent 1 -Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL: RG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 3, 4)Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RSI.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All Definition(s): SITE BOUNDARY-Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan BNP Basis: Dose assessments are performed by computer-based methods (ref. 1, 2, 3)NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.I OPEP-02.2.1 Rev. 6 1 Page 53 of303 ATTACHMENT 1 Page 28 of 210 EAL Bases 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.BNP Basis Reference(s):

1. EMG-NGGC-0002 Off-site Dose Assessment

2. OPEP-03.4.7 Automation of Off-Site Dose Projections

3. 0E&RC-03.4.8, Offsite Dose Projections for Monitored Releases 4. NEI 99-01 AG1 OPEP-02.2.1 Rev. 6 Page 54 of 303]

ATTACHMENT 1 Page 29 of 210 EAL Bases Category: Subcategory: Initiating Condition: R -Abnormal Rad Levels I Rad Effluent 1 -Radiological Effluent Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL: RG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:* Closed window dose rates > 1,000 mR/hr expected to continue for > 60 min." Analyses of field survey samples indicate thyroid CDE > 5,000 mrem for 60 min. of inhalation.(Notes 1,2)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Mode Applicability: All Definition(s): SITE BOUNDARY-Area as depicted in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan BNP Basis: OPEP-02.6.6, Environmental Monitoring Team Leader and OPEP-03.5.5 Environmental Monitoring and Plume Tracking provide guidance for emergency or post-accident radiological environmental monitoring (ref. 1, 2).OPEP-02.2.1 Rev. 6 Page 55 of 303] ATTACHMENT 1 Page 30 of 210 EAL Bases NEI 99-01 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.BNP Basis Reference(s):

1. 0PEP-02.6.6, Environmental Monitoring Team Leader 2. 0PEP-03.5.5 Environmental Monitoring and Plume Tracking 3. NEI 99-01 AG1 OPEP-02.2.1 Rev. 6 Page 56 of 303]

ATTACHMENT 1 Page 31 of 210 EAL Bases Category: Subcategory: R -Abnormal Rad Levels I Rad Effluent 2 -Irradiated Fuel Event Initiating Condition: Unplanned loss of water level above irradiated fuel EAL: RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm (A-04 6-6) or indication AND UNPLANNED rise in area radiation levels as indicated by any of the following radiation monitors: " ARM Channel 26 New Fuel Vault" ARM Channel 27 North of Fuel Pool* ARM Channel 28 Between Reactor and Fuel Pool 0 ARM Channel 29 Cask Wash Area Mode Applicability: All Definition(s): 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.REFUELING PATHWAY-. The reactor refueling cavity, spent fuel pool and fuel transfer canal (cattle chute) comprise the refueling pathway.I OPEP-02.2.1 I Rev. 6 Page 57 of 303 ATTACHMENT 1 Page 32 of 210 EAL Bases BNP Basis: The spent fuel pool low water level alarm setpoint is actuated by level switch G410-LSHL-NO01 at a setpoint of 37' 5". Water level restoration instructions are performed in accordance with 1(2)APP A-04 6-6 Fuel Pool Level Low (ref. 1).When the spent fuel pool and reactor cavity are connected, there could exist the possibility of uncovering irradiated fuel. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred to and from the RPV and spent fuel pool.NEI 99-01 Basis: 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.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.BNP Basis Reference(s):

1. 1(2)APP-A-04 6-6 Fuel Pool Level Low 2. DBD-1 1 Radiation Monitoring System 3. NEI 99-01 AU2 I OPEP-02.2.1 Rev. 6 1 Page 58 of 303 ATTACHMENT 1 Page 33 of 210 EAL Bases Category: Subcategory:

R -Abnormal Rad Levels / Rad Effluent 2 -Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: RA2.1 Unusual Event Uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability: All Definition(s): REFUELING PATHWAY-. The reactor refueling cavity, spent fuel pool and fuel transfer canal (cattle chute) comprise the refueling pathway.BNP Basis: None.NEI 99-01 Basis: This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. 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.I OPEP-02.2.1 Rev. 6 1 Page 59 of 303 ATTACHMENT 1 Page 34 of 210 EAL Bases This EAL escalates from RU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations. While an area radiation monitor could detect an 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.Escalation of the emergency classification level would be via IC RS1.BNP Basis Reference(s):

1. 1(2)APP-A4 6-6 (Fuel pool Level Low)2. 1(2)APP-A7 2-2 (Reactor Water Level Hi/Low)3. NEI 99-01 AA2 I OPEP-02.2.1 Rev. 6 1 Page 60 of 303 ATTACHMENT 1 Page 35 of 210 EAL Bases Category: Subcategory:

R -Abnormal Rad Levels / Rad Effluent 2 -Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND Any of the following radiation monitor indications: " Reactor Bldg Vent Rad Monitor Channel A or B (> 3 mR/hr)" ARM Channel 26 New Fuel Vault (> 6 mR/hr)* ARM Channel 27 North of Fuel Pool (>10 mR/hr)" ARM Channel 28 Between Reactor and Fuel Pool (> 1000 mR/hr)" ARM Channel 29 Cask Wash Area (>40 mR/hr)Mode Applicability: All Definition(s): None BNP Basis: The high alarm setpoints for the radiation monitors are (ref. 1, 2, 3 4):* Reactor Building Exhaust Plenum Rad Monitor Channel A or B > 3 mR/hr" ARM Channel 26 New Fuel Vault > 6 mR/hr" ARM Channel 27 North of Fuel Pool > 10 mR/hr" ARM Channel 28 Between Reactor and Fuel Pool > 1000 mR/hr" ARM Channel 29 Cask Wash Area > 40mR/hr OPEP-02.2.1 Rev. 6 Page 61 of 303] ATTACHMENT 1 Page 36 of 210 EAL Bases NEI 99-01 Basis: This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. 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.Escalation of the emergency would be based on either Recognition Category R or C ICs.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). Escalation of the emergency classification level would be via IC RSI.BNP Basis Reference(s):

1. 1(2)APP-UA-03 3-7 2. 1(2)APP-UA-03 4-5 3. 1(2)APP-UA-03 4-7 4. DBD-1 1 Radiation Monitoring System 5. NEI 99-01 AA2 I OPEP-02.2.1 Rev. 6 Page 62 of 303 ATTACHMENT 1 Page 37 of 210 EAL Bases Category: Subcategory:

R -Abnormal Rad Levels / Rad Effluent 2 -Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: RA2.3 Alert Lowering of spent fuel pool level to < 105 ft. 3 in. ele.Mode Applicability: All Definition(s): None BNP Basis: Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1 -116 ft. 1 in. ele.), SFP level 10 ft. above the top of the fuel racks (Level 2 -105 ft. 3 in. ele.) and SFP level at the top of the fuel racks (Level 3 -95 ft. 3 in. ele.).An indicated level of 105 ft. 3 in. corresponds to the Level 2 setpoint (Ref. 1).NEI 99-01 Basis: This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. 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.Escalation of the emergency would be based on either Recognition Category R or C ICs.OPEP-02.2.1 Rev. 6 Page 63 of 303] ATTACHMENT 1 Page 38 of 210 EAL Bases 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 IC RSI.BNP Basis Reference(s):

1. PCHG-DESG Engineering Change 0000089578R0

2. NEI 99-01 AA2 OPEP-02.2.1 Rev. 6 Page 64 of 303]

ATTACHMENT 1 Page 39 of 210 EAL Bases Category: Subcategory: R -Abnormal Rad Levels / Rad Effluent 2 -Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL: RS2.1 Site Area Emergency Lowering of spent fuel pool level to < 95 ft. 3 in. ele.Mode Applicability: All Definition(s): None BNP Basis: Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1 -116 ft. 1 in. ele.), SFP level 10 ft. above the top of the fuel racks (Level 2 -105 ft. 3 in. ele.) and SFP level at the top of the fuel racks (Level 3 -95 ft. 3 in. ele.).An indicated level of 95 ft. 3 in. corresponds to the Level 3 setpoint (Ref. 1).NEI 99-01 Basis: This EAL 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 AG1 or RG2.OPEP-02.2.1 Rev. 6 Page 65 of 303] ATTACHMENT 1 Page 40 of 210 EAL Bases BNP Basis Reference(s):

1. PCHG-DESG Engineering Change 0000089578R0

2. NEI 99-01 AS2 I OPEP-02.2.1 I Rev. 6 Page 66 of 3031 ATTACHMENT 1 Page 41 of 210 EAL Bases Category: Subcategory:

Initiating Condition: R -Abnormal Rad Levels / Rad Effluent 2 -Irradiated Fuel Event Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL: RG2.1 General Emergency Spent fuel pool level cannot be restored > 95 ft. 3 in. ele. for_> 60 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: All Definition(s): None BNP Basis: Post-Fukushima order EA-1 2-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1 -116 ft. 1 in. ele.), SFP level 10 ft. above the top of the fuel racks (Level 2 -105 ft. 3 in. ele.) and SFP level at the top of the fuel racks (Level 3 -95 ft. 3 in. ele.).An indicated level of 95 ft. 3 in. corresponds to the Level 3 setpoint (Ref. 1).NEI 99-01 Basis: This EAL 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. OPEP-02.2."I Rev. 6 Page 67 of 303] ATTACHMENT 1 Page 42 of 210 EAL Bases BNP Basis Reference(s):

1. PCHG-DESG Engineering Change 0000089578R0

2. NEI 99-01 AG2 OPEP-02.2.1 Rev. 6 Page 68 of 303]

ATTACHMENT 1 Page 43 of 210 EAL Bases Category: Subcategory: R -Abnormal Rad Levels / Rad Effluent 3- Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL: RA3.1 Alert Dose rates > 15 mR/hr in EITHER of the following areas: Control Room (ARM Channel 1-1)OR Central Alarm Station (by survey)Mode Applicability: All Definition(s): None BNP Basis: Areas that meet this threshold include the Control Room and the Central Alarm Station (CAS).ARM Channel 1-1 monitors the Control room for area radiation (ref. 1). The CAS is included in this EAL because of its' importance to permitting access to areas required to assure safe plant operations. There is no permanently installed CAS area radiation monitors that may be used to assess this EAL threshold. Therefore this threshold must be assessed via local radiation survey for the CAS.OPEP-02.2.1 Rev. 6 Page 69 of 303] ATTACHMENT 1 Page 44 of 210 EAL Bases NEI 99-01 Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Site Emergency Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable. Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.BNP Basis Reference(s):

1. 1(2)APP-UA-03 6-7 (Area RAD Control Room Hi)2. NEI 99-01 AA3 I OPEP-02.2.1 Rev. 6 1 Page 70 of 303 1 ATTACHMENT 1 Page 45 of 210 EAL Bases Category: Subcategory:

R -Abnormal Rad Levels / Rad Effluent 3 -Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL: RA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table R-2 rooms or areas (Note 5)Note 5: 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. Table R-2 Safe Operation & Shutdown Areas RoomlArea Mode Applicability Reactor Building -17' North RHR Unit-1 & 2 3, 4, 5 Reactor Building -17' South RHR Unit-1 & 2 3, 4, 5 Reactor Building 20' East & West MCC Areas Unit-1 & 2 3, 4, 5 Reactor Building 20' Pipe Tunnel Unit-1 & 2 3, 4, 5 Mode Applicability: All Definition(s): IMPEDE(D)- Personnel access to a room or area is hindered to an extent that 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). 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.BNP Basis: If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.OPEP-02.2.1 Rev. 6 Page 71 of 303 ATTACHMENT 1 Page 46 of 210 EAL Bases The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).NEI 99-01 Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Site Emergency Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable. For 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: " 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.OPEP-02.2.1 Rev. 6 Page 72 of 303] ATTACHMENT 1 Page 47 of 210 EAL Bases* If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.BNP Basis Reference(s):

1. Attachment 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases 2. NEI 99-01 AA3 I OPEP-02.2.1 1 Rev. 6 Page 73 of 303 1 ATTACHMENT 1 Page 48 of 210 EAL Bases Category C -Cold Shutdown I Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature < 200*F); EALs in this category are applicable only in one or more cold operating modes.Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, containment closure, and fuel clad integrity for the applicable operating modes (4 -Cold Shutdown, 5 -Refueling, D -Defueled). The events of this category pertain to the following subcategories:

1. RPV Level Reactor Pressure Vessel water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.

2. Loss of Emergency AC Power Loss of essential plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

This category includes loss of onsite and offsite power sources for 4160 VAC emergency buses.3. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.

4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

This category includes loss of power to or degraded voltage on the 125 VDC vital buses.0PEP-02.2.1 Rev. 6 Page 74 of 303] ATTACHMENT 1 Page 49 of 210 EAL Bases 5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

6. Hazardous Event Affectinq Safety Systems Certain hazardous natural and technological events may result in visible damage to or degraded performance of safety systems warranting classification.

OPEP-02.2.1 Rev. 6 Page 75 of 303] ATTACHMENT 1 Page 50 of 210 EAL Bases Category: Subcategory: C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Initiating Condition: UNPLANNED loss of RPV inventory for 15 minutes or longer EAL: CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RPV water level less than a required lower limit for > 15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): 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.BNP Basis: Figure C-1 illustrates the elevations of the RPV level instrument ranges (ref. 2).With the plant in Cold Shutdown, RPV water level is normally maintained above the RPV low level scram setpoint of 166 in. above TAF (ref. 1, 3). However, if RPV level is being controlled below the RPV low level scram setpoint, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern.With the plant in Refueling mode, RPV water level is normally maintained at or above the reactor vessel flange (Technical Specification LCO 3.9.6 requires at least 23 ft of water above the top of the reactor vessel flange in the refueling cavity during refueling operations). The RPV flange is at an indicated level of 355 in. as indicated on the red scale of B21-LI-R605A/B Shutdown Range Reactor Water Level Indication (ref. 4).0PEP-02.2.1 Rev. 6 Page 76 of 3031 ATTACHMENT 1 Page 51 of 210 EAL Bases NEI 99-01 Basis: This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor 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.This EAL 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.Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.BNP Basis Reference(s):

1. OEOP-01-NL EOP-SAMG NUMERICAL LIMITS AND VALUES, Table 1E 2. SD-01.2 Reactor Vessel Instrumentation Figure 01.2-1 Reactor Water Level Instrument Ranges 3. 1(2) APP A7 2-2 (Reactor Water Level Hi/Low)4. OGP-06 Cold Shutdown to Refueling (Head Unbolted) step 5.1.14 5. NEI 99-01 CU1 I OPEP-02.2.1 Rev. 6 1 Page 77 of 303 1 ATTACHMENT 1 Page 52 of 210 EAL Bases Figure C-1 RPV Levels (ref. 2)Reactor Water Level Instrument Ranges 917" A FROM VESSEL ZERO Y- YI=06PPROX 850'I -[I_1]HML STEAM UNE SPILL-OVER 4 IS 254k ON N027 5 H U T D 0 w N R A N G E NARROW RANGE H0r-NORMAL LEVEL -f w I D E R A N G E N027 (-8.44")LCOE F U E L z 0 N E N026 mt -.1 hi NOSSJN037 1\1, r I OPEP-02.2.1 Rev. 6 Page 78 of 303 1 ATTACHMENT 1 Page 53 of 210 EAL Bases Category: Subcategory:

C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Initiating Condition: UNPLANNED loss of RPV inventory EAL: CUl.2 Unusual Event RPV water level cannot be monitored AND UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory Table C-1 Sumps & Tanks ]* Drywell Floor Drain Sump" Drywell Equipment Drain Sump* RB Floor Drain Sump* RB Equipment Drain Sump* Torus* Visual Observation Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): 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.IOPEP-02.2.1 I Rev. 6 Page 79 of 303 ATTACHMENT 1 Page 54 of 210 EAL Bases BNP Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refuel mode is normally monitored using the red scale of B21-LI-R605A/B Shutdown Range Reactor Water Level Indication. In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (ref. 1). A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage (ref. 5). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. NEI 99-01 Basis: This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor 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.This EAL 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 RCS.Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.I OPEP-02.2.1 Rev. 6 1 Page 80 of 303 1 ATTACHMENT 1 Page 55 of 210 EAL Bases BNP Basis Reference(s):

1. OOP-47 Floor and Equipment Drain System Operating Procedure 2. 101-03.1 Control Room Operator Daily Surveillance Report 3. 201-03.2 Control Room Operator Daily Surveillance Report 4. OAOP-14.0 Abnormal Primary Containment Conditions

5. 1(2)OP-17 Residual Heat Removal System Operating Procedure 6. NEI 99-01 CU1 OPEP-02.2.1 Rev. 6 Page 81 of 303]

ATTACHMENT 1 Page 56 of 210 EAL Bases Category: Subcategory: C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Initiating Condition: Loss of RPV inventory EAL: CA1.1 Alert Loss of RPV inventory as indicated by RPV water level < 105 in. above TAF (Level 2)Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): None BNP Basis: The threshold RPV level of 105 in. is the low-low ECCS actuation setpoint (ref. 1). RPV level is normally monitored using the instruments in Figure C-1 (ref. 2).When reactor vessel water level drops to 105 in. above TAF high pressure steam-driven injection sources HPCI (ECCS) and RCIC receive an initiation signal (ref. 1). Although these systems cannot restore RCS inventory in the cold condition, the Low-Low (Level 2) ECCS actuation setpoint is operationally significant and is indicative of a loss of RCS inventory significantly below the low level scram setpoint specified in CUI.1.NEI 99-01 Basis: This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.For this EAL, a lowering of water level below 105 in above TAF 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.I OPEP-02.2.1 Rev. 6 1 Page 82 of 303 1 ATTACHMENT 1 Page 57 of 210 EAL Bases Although related, this EAL 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 Decay Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.If RPV water level continues to lower, then escalation to Site Area Emergency would be via IC CSI.BNP Basis Reference(s):

1. OEOP-01-NL EOP-SAMG NUMERICAL LIMITS AND VALUES, Table 1E 2. SD-011.2 Reactor Vessel Instrumentation Figure 01.2-1 Reactor Water Level Instrument Ranges OPEP-02.2.1 Rev. 6 Page 83 of 303]

ATTACHMENT 1 Page 58 of 210 EAL Bases Figure C-1 RPV Levels (ref. 2)Reactor Water Level Instrument Ranges 917" , APPROX 850' FROM VESSEL ZERO APPO 150 OPEP-02.2.1 Rev. 6 Page 84 of 303] ATTACHMENT 1 Page 59 of 210 EAL Bases Category: Subcategory: C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Initiating Condition: Loss of RPV inventory EAL: CA1.2 Alert RPV water level cannot be monitored for > 15 min. (Note 1)AND UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory Table C-1 Sumps & Tanks* Drywell Floor Drain Sump* Drywell Equipment Drain Sump* RB Floor Drain Sump" RB Equipment Drain Sump" Torus* Visual Observation Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): 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.BNP Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. In the Refuel mode, the RCS is not intact and RPV level may be monitored by different means, including the ability to monitor level visually.OPEP-02.2.1 Rev. 6 Page 85 of 303] ATTACHMENT 1 Page 60 of 210 EAL Bases In this EAL, all water level indication would be unavailable, and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Sump level increases must be evaluated against other potential sources of leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (ref. 1 thru 4). A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV.With RHR System operating in the Shutdown Cooling mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage (ref. 5). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. NEI 99-01 Basis: This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.For this EAL, 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 RCS.The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1.If the RCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.OPEP-02.2.1 Rev. 6 Page 86 of 303] ATTACHMENT 1 Page 61 of 210 EAL Bases BNP Basis Reference(s): 1.2.3.4.5.6.0OP-47 Floor and Equipment Drain System Operating Procedure 101-03.1 Control Room Operator Daily Surveillance Report 201-03.2 Control Room Operator Daily Surveillance Report OAOP-14.0 Abnormal Primary Containment Conditions 1(2)OP-17 Residual Heat Removal System Operating Procedure NEI 99-01 CA1 I OPEP-02.2.1 I Rev. 6 Page 87 of 303 1 ATTACHMENT 1 Page 62 of 210 EAL Bases Category: Subcategory: C -Cold Shutdown I Refueling System Malfunction 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1.1 Site Area Emergency CONTAINMENT CLOSURE not established AND RPV level < 45 in. (Level 3)Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): CONTAINMENT CLOSURE -The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to BNP, Containment Closure is established when either Primary Containment is Operable per Section 3.6.1.1 of Technical Specifications or Secondary Containment is considered functional per the requirements of OAP-022, BNP Outage Risk Management. BNP Basis: RPV level is normally monitored using the instruments in Figure C-1 (ref. 2).When RPV level decreases to 45 in., RPV water level is below the low-low-low ECCS actuation setpoint (Level 3) (ref. 1).The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV water level decrease and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier.OPEP-02.2.1 Rev. 6 Page 88 of 303] ATTACHMENT 1 Page 63 of 210 EAL Bases NEI 99-01 Basis: This IC addresses a significant and prolonged loss of RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant.These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable.Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RPV levels of CS1.1 and CS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. This EAL addresses 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 CG1 or RG1.BNP Basis Reference(s):

1. OEOP-01-NL EOP-SAMG NUMERICAL LIMITS AND VALUES, Table 1E 2. BNP Technical Specifications, Sections 3.6.1.1 3. OAP-022, BNP Outage Risk Management, Section 6.5 4. SD-011.2 Reactor Vessel Instrumentation Figure 01.2-1 Reactor Water Level Instrument Ranges 5. NEI 99-01 CS1 I OPEP-02.2.1 Rev. 6 Page 89 of 303]

ATTACHMENT 1 Page 64 of 210 EAL Bases Figure C-1 RPV Levels (ref. 4)Reactor Water Level Instrument Ranges 917" 4 FROM VESSEL ZERO kPPROX 850'r OPEP-02.2.1 Rev. 6 Page 90 of 303] ATTACHMENT 1 Page 65 of 210 EAL Bases Category: Subcategory: C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1.2 Site Area Emergency CONTAINMENT CLOSURE established AND RPV level < TAF Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): CONTAINMENT CLOSURE -The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to BNP, Containment Closure is established when either Primary Containment is Operable per Section 3.6.1.1 of Technical Specifications or Secondary Containment is considered functional per the requirements of OAP-022, BNP Outage Risk Management. BNP Basis: When RPV level drops to the top of active fuel (TAF) (an indicated RPV level of --8 in.), core uncovery starts to occur (ref. 1).IOPEP-02.2.1 I Rev. 6 Page 91 of 303 ATTACHMENT 1 Page 66 of 210 EAL Bases NEI 99-01 Basis: This IC addresses a significant and prolonged loss of RPV level 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 RPV level cannot be restored, fuel damage is probable.Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of CS1.1 and CS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. This EAL addresses 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 CG1 or RG1.BNP Basis Reference(s):

1. OEOP-01-NL EOP-SAMG NUMERICAL LIMITS AND VALUES 2. BNP Technical Specifications, Sections 3.6.1.1 and 3.6.4.1 3. OAP-022, BNP Outage Risk Management, Section 6.5 IOPEP-02.2.1 Rev. 6 Page 92 of 303]

ATTACHMENT 1 Page 67 of 210 EAL Bases Category: Subcategory: C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1.3 Site Area Emergency RPV water level cannot be monitored for -> 30 min. (Note 1)AND Core uncovery is indicated by EITHER of the following: " UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory* UNPLANNED increase in ARM Channel 28 Between Reactor and Fuel Pool> 1000 mR/hr Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table C-1 Sumps & Tanks" Drywell Floor Drain Sump" Drywell Equipment Drain Sump* RB Floor Drain Sump" RB Equipment Drain Sump" Torus" Visual Observation Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): 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.OPEP-02.2.1 Rev. 6 Page 93 of 303] ATTACHMENT 1 Page 68 of 210 EAL Bases BNP Basis: If RPV level monitoring capability is unavailable, the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Sump level increases must be evaluated against other potential sources of leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (ref. 1 thru 4). A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage (ref. 5). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory. In the Refueling Mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in indications on installed area radiation monitors. ARM Channel 28 Between Reactor and Fuel Pool is located on the Refuel Floor in the Reactor Building and is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations. If this radiation monitor reaches and exceeds the alarm setpoint of 1,000 mr/hr, a loss of inventory with potential to uncover the core is likely to have occurred (Ref. 6, 7).NEI 99-01 Basis: This IC addresses a significant and prolonged loss of reactor vessel/RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant.These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable.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. I OPEP-02.2.1 Rev. 6 Page 94 of 303] ATTACHMENT 1 Page 69 of 210 EAL Bases 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 RCS.This EAL addresses 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 CG1 or RG1 BNP Basis Reference(s):

1. OOP-47 Floor and Equipment Drain System Operating Procedure 2. 101-03.1 Control Room Operator Daily Surveillance Report 3. 201-03.2 Control Room Operator Daily Surveillance Report 4. OAOP-14.0 Abnormal Primary Containment Conditions

5. 1(2)OP-17 Residual Heat Removal System Operating Procedure 6. 1(2)APP-UA-03 4-7 7. DBD-1 1 Radiation Monitoring System 8. NEI 99-01 CS1 OPEP-02.2.1 Rev. 6 Page 95 of 303]

ATTACHMENT 1 Page 70 of 210 EAL Bases Category: Subcategory: Initiating Condition: C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Loss of RPV inventory affecting fuel clad integrity with Containment challenged EAL: CGI.1 General Emergency RPV level < TAF for - 30 min. (Note 1)AND Any Containment Challenge indication, Table C-2 Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.Table C-2 Containment Challenge Indications

CONTAINMENT CLOSURE not established (Note 6)* Primary Containment hydrogen concentration

> 6%* UNPLANNED rise in PC pressure* Exceeding one or more Secondary Containment Control Maximum Safe Operating Area Radiation Levels (0EOP-03-SCCP Table 3)Mode Applicability: 4 -Cold Shutdown, 5- Refuel I 0PEP-02.2.1 Rev. 6 1 Page 96 of 303 1 ATTACHMENT 1 Page 71 of 210 EAL Bases Definition(s): CONTAINMENT CLOSURE -The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to BNP, Containment Closure is established when either Primary Containment is Operable per Section 3.6.1.1 of Technical Specifications or Secondary Containment is considered functional per the requirements of OAP-022, BNP Outage Risk Management. BNP Basis: When RPV level drops below --8 in., the top of active fuel, core uncovery starts to occur (ref.6).Four conditions are associated with a challenge to Primary Containment (PC) integrity: " CONTAINMENT CLOSURE is not established (Ref. 7).* 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 mixture of dissolved gases in the Primary Containment. However, Primary Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. An explosive mixture can be formed when hydrogen gas concentration in the Primary Containment atmosphere is greater than 6%by volume in the presence of oxygen (>5%) (ref. 2, 3). In Cold Shutdown and Refueling modes it is assumed that the Primary Containment is de-inerted. Monitors CAC-AT-4409 and 4410 monitor hydrogen and oxygen concentration in the drywell during normal operation. These monitors are isolated by accident isolation signals. However, monitors CAC-AT-4409 and 4410 will be realigned to the primary containment for post-accident monitoring via an operator actuated isolation signal override circuit when directed by the EOPs. (ref. 4)." Any unplanned increase in PC pressure in the Cold Shutdown or Refueling mode indicates a potential loss of containment closure capability. Unplanned Primary Containment pressure increases indicates containment closure cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release." Secondary Containment radiation monitors should provide indication of increased release that may be indicative of a challenge to Containment Closure. The Maximum Safe Operating Values are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in OEOP-03-SCCP, Secondary Containment Control, (ref. 5).I OPEP-02.2.1 I Rev. 6 Page 97 of 303 ATTACHMENT 1 Page 72 of 210 EAL Bases NEI 99-01 Basis: 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 RPV level cannot be restored, fuel damage is probable.With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. 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. This EAL addresses 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. OPEP-02.2.1 Rev. 6 Page 98 of 303] ATTACHMENT 1 Page 73 of 210 EAL Bases BNP Basis Reference(s): 1.2.3.4.5.6.7.8.BNP Technical Specifications Sections 3.6.1.1 and 3.6.4.1 BWROG EPG/SAG Revision 2, Sections PC/G OEOP-02-PCCP, Primary Containment Control Updated FSAR section 6.2.5.2.2 OEOP-03-SCCP, Secondary Containment OEOP-01-NL EOP-SAMG NUMERICAL LIMITS AND VALUES OAP-022 BNP Outage Risk Management, Section 6.5 NEI 99-01 CG1 OPEP-02.2.1 Rev. 6 Page 99 of 303] ATTACHMENT 1 Page 74 of 210 EAL Bases Category: Subcategory: Initiating Condition: EAL: C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Loss of RPV inventory affecting fuel clad integrity with Containment challenged CG1.2 General Emergency RPV level cannot be monitored for -> 30 min. (Note 1)AND Core uncovery is indicated by EITHER of the following:

UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory* UNPLANNED increase in ARM Channel 28 Between Reactor and Fuel Pool> 1000 mR/hr AND Any Containment Challenge indication, Table C-2 Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.Table C-1 Sumps & Tanks" Drywell Floor Drain Sump" Drywell Equipment Drain Sump* RB Floor Drain Sump" RB Equipment Drain Sump" Torus" Visual Observation IOPEP-02.2." I Rev. 6 1 Page 100 of303 ATTACHMENT 1 Page 75 of 210 EAL Bases Table C-2 Containment Challenge Indications
CONTAINMENT CLOSURE not established (Note 6)* Primary Containment hydrogen concentration

> 6%* UNPLANNED rise in PC pressure* Exceeding one or more Secondary Containment Control Maximum Safe Operating Area Radiation Levels (OEOP-03-SCCP Table 3)Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): CONTAINMENT CLOSURE -The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to BNP, Containment Closure is established when either Primary Containment is Operable per Section 3.6.1.1 of Technical Specifications or Secondary Containment is considered functional per the requirements of OAP-022, BNP Outage Risk Management. 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.BNP Basis: If RPV level monitoring capability is unavailable, the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Sump level increases must be evaluated against other potential sources of leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (ref. 1 thru 4). A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage (ref. 5). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. OPEP-02.2.1 Rev. 6 Page 101 of 303 ATTACHMENT 1 Page 76 of 210 EAL Bases In the Refueling Mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in indications on installed area radiation monitors. ARM Channel 28 Between Reactor and Fuel Pool is located on the Refuel Floor in the Reactor Building and is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations. If this radiation monitor reaches and exceeds the alarm setpoint of 1,000 mr/hr, a loss of inventory with potential to uncover the core is likely to have occurred (Ref. 6, 7).Four conditions are associated with a challenge to Primary Containment (PC) integrity: " CONTAINMENT CLOSURE is not established (Ref. 13).* 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 mixture of dissolved gases in the Primary Containment. However, Primary Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. An explosive mixture can be formed when hydrogen gas concentration in the Primary Containment atmosphere is greater than 6%by volume in the presence of oxygen (>5%) (ref. 9, 10). In Cold Shutdown and Refueling modes it is assumed that the Primary Containment is de-inerted. Monitors CAC-AT-4409 and 4410 monitor hydrogen and oxygen concentration in the drywell during normal operation. These monitors are isolated by accident isolation signals. However, monitors CAC-AT-4409 and 4410 will be realigned to the primary containment for post-accident monitoring via an operator actuated isolation signal override circuit when directed by the EOPs. (ref. 11).* Any unplanned increase in PC pressure in the Cold Shutdown or Refueling mode indicates a potential loss of containment closure capability. Unplanned Primary Containment pressure increases indicates containment closure cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release.* Secondary Containment radiation monitors should provide indication of increased release that may be indicative of a challenge to Containment Closure. The Maximum Safe Operating Values are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in OEOP-03-SCCP, Secondary Containment Control, (ref. 12).I OPEP-02.2.1 I Rev. 6 Page 102 of 303 ATTACHMENT 1 Page 77 of 210 EAL Bases NEI 99-01 Basis: 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 RPV level cannot be restored, fuel damage is probable.With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. 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 RCS.I OPEP-02.2.1 Rev. 6 1 Page 103 of 303 ATTACHMENT 1 Page 78 of 210 EAL Bases This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal;SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1 449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. BNP Basis Reference(s):

1. OOP-47 Floor and Equipment Drain System Operating Procedure 2. 101-03.1 Control Room Operator Daily Surveillance Report 3. 201-03.2 Control Room Operator Daily Surveillance Report 4. OAOP-14.0 Abnormal Primary Containment Conditions

5. 1(2)OP-17 Residual Heat Removal System Operating Procedure 6. 1(2)APP-UA-03 4-7 7. DBD-1 1 Radiation Monitoring System 8. BNP Technical Specifications Section 3.6.1.1 and 3.6.4.1 9. BWROG EPG/SAG Revision 2, Sections PC/G 10. OEOP-02-PCCP Primary Containment Control 11. Updated FSAR section 6.2.5.2.2 12. OEOP-03-SCCP Secondary Containment Control 13. 0AP-022 BNP Outage Risk Management, Section 6.5 14. NEI 99-01 CG1 I OPEP-02.2.1 I Rev. 6 Page 104 of 303 ATTACHMENT 1 Page 79 of 210 EAL Bases Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 2 -Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer EAL: CU2.1 Unusual Event AC power capability to Emergency 4 KV Buses El (E3) and E2(E4) reduced to a single power source for > 15 min. (Note 1)AND Any additional single power source failure will result in loss of all unit-specific AC power to SAFETY SYSTEMS Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 4 -Cold Shutdown, 5 -Refuel, D -Defueled Definition(s): 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 (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. IOPEP-02.2.1 I Rev. 6 Page10 of0303 ATTACHMENT 1 Page 80 of 210 EAL Bases BNP Basis: The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. 4160 VAC power is divided into Balance of Plant (BOP Bus), and Emergency (E-Bus) distribution. The BOP Buses consist of the Conventional Buses (Common A/B Buses;Buses 1 B/2B, 1 C/2C, 1 D/2D) and Caswell Beach Buses A/B. The Emergency switchgear are Buses El/E3 (Division I) and E2/E4 (Division II) (Ref. 1,2).The E-Buses are normally powered through the respective BOP Buses (1 D to El, 1C to E2, 2D to E3, 2C to E4) via a master/slave breaker arrangement. Each E-Bus has a dedicated Diesel Generator to supply an emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of offsite power. The DGs will automatically start and tie onto the E-Buses if the normal power source or offsite power is lost. The DGs can provide power to the E-Buses only. In the event the diesel generator is unavailable for an E-Bus, crosstie capability exists for each E-Bus from the same division of the opposite unit (El to E3, E2 to E4). Although the E-Buses within the unit also have crosstie capability, this alignment is not permitted by plant procedures, with the exception of El to E2 during specific Alternate Safe Shutdown (ASSD) conditions. During periods of unit shutdown, when the Startup Auxiliary Transformer (SAT) would be the only normal source of offsite power, the Unit Auxiliary Transformer (UAT) can be made available by establishing a UAT backfeed. Backfeed from the UAT will require the use of keys for the control selector switches and opening of the respective generator's manual no-load disconnect (Ref. 3).Because 2 RHR pumps on each unit are powered from the unaffected unit, the words "unit-specific" have been added to clarify that the cross-connected RHR pump power cannot be credited as an AC power source relative to this EAL.This cold condition EAL is equivalent to the hot condition EAL SA1.1.NEI 99-01 Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as 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.OPEP-02.2.1 Rev. 6 Page 106 of 303 ATTACHMENT 1 Page 81 of 210 EAL Bases An "AC power source" is a source recognized in AOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below." A loss of all offsite power with a concurrent failure of one division of emergency power sources (e.g., onsite diesel generators)." A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator." A loss of emergency power sources (e.g., onsite diesel generators) with a single division of emergency 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.BNP Basis Reference(s): 1.2.3.4.5.6.7.8.Drawing BN-50.0.01 Electrical Distribution BNP Updated FSAR Chapter 8 1(2)OP-50 Plant Electric System Operating 1(2)APP-UA15 2-1 (Bus E-1 Undervoltage) 1(2)APP-UA16 2-1 (Bus E-2 Undervoltage) 1(2)APP-UA17 2-1 (Bus E-3 Undervoltage) 1(2)APP-UA18 2-1 (Bus E-4 Undervoltage) NEI 99-01 CU2 Procedure I OPEP-02.2.1 I Rev. 6 1 Page 107 of 303 ATTACHMENT 1 Page 82 of 210 EAL Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 2 -Loss of Emergency AC Power Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer EAL: CA2.1 Alert Loss of all offsite and all onsite AC power capability to Emergency 4 KV Buses El (E3)and E2(E4) for _ 15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 4 -Cold Shutdown, 5 -Refuel, D -Defueled BNP Basis: The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. 4160 VAC power is divided into Balance of Plant (BOP Bus), and Emergency (E-Bus) distribution. The BOP Buses consist of the Conventional Buses (Common A/B Buses;Buses 1 B/2B, 1 C/2C, 1 D/2D) and Caswell Beach Buses A/B. The Emergency switchgear are Buses El/E3 (Division I) and E2/E4 (Division II) (Ref. 1,2).The E-Buses are normally powered through the respective BOP Buses (1D to El, IC to E2, 2D to E3, 2C to E4) via a master/slave breaker arrangement. Each E-Bus has a dedicated Diesel Generator to supply an emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of offsite power. The DGs will automatically start and tie onto the E-Buses if the normal power source or offsite power is lost. The DGs can provide power to the E-Buses only. In the event the diesel generator is unavailable for an E-Bus, crosstie capability exists for each E-Bus from the same division of the opposite unit (El to E3, E2 to E4). Although the E-Buses within the unit also have crosstie capability, this alignment is not permitted by plant procedures, with the exception of El to E2 during specific Alternate Safe Shutdown (ASSD) conditions. During periods of unit shutdown, when the Startup Auxiliary Transformer (SAT) would be the only normal source of offsite power, the Unit Auxiliary Transformer (UAT) can be made available by establishing a UAT backfeed. Backfeed from the UAT will require the use of keys for the control selector switches and opening of the respective generator's manual no-load disconnect (Ref. 3).OPEP-02.2.1 Rev. 6 Page 108 of 303 ATTACHMENT 1 Page 83 of 210 EAL Bases This cold condition EAL is equivalent to SS1.1.the hot condition loss of all offsite AC power EAL NEI 99-01 Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level would be via IC CS1 or RS1.BNP Basis Reference(s): 1.2.3.4.5.6.7.8.Drawing BN-50.0.01 Electrical Distribution BNP Updated FSAR Chapter 8 1(2)OP-50 Plant Electric System Operating Procedure 1(2)APP-UA15 2-1 (Bus E-1 Undervoltage) 1(2)APP-UA16 2-1 (Bus E-2 Undervoltage) 1(2)APP-UA17 2-1 (Bus E-3 Undervoltage) 1(2)APP-UA18 2-1 (Bus E-4 Undervoltage) NEI 99-01 CA2 I OPEP-02.2.1 I Rev. 6 Page 109 of 303 ATTACHMENT 1 Page 84 of 210 EAL Bases Category: Subcategory: C -Cold Shutdown / Refueling System Malfunction 3 -RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL: CU3.1 Unusual Event UNPLANNED increase in RCS temperature to > 212°F due to loss of decay heat removal capability Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): 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.BNP Basis: Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (212 0 F, ref. 1). These include (ref. 2):* Recirculation Suction Temperatures read on B32-TR-R650 located on panel P-603 (if recirculation loop is in operation)

RHR HX 2A(B) Inlet Temperature as read on E41-TR-R605 Point 1(2), on Panel H12-P614 (RHR HX in service)" RHR HX 2A(B) Outlet Temperature as read on E41-TR-R605 Point 3(4), on Panel H12-P614 (RHR HX not in service)* PPC Display 815, RPV HEATUP/COOLDOWN MONITOR (natural circulation)

NEI 99-01 Basis: This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limitand represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Site Emergency Coordinator should also refer to IC CA3.I OPEP-02.2.1 Rev. 6 Page 110 of303 ATTACHMENT 1 Page 85 of 210 EAL Bases A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. This EAL 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 at or 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.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.BNP Basis Reference(s):

1. BNP Technical Specifications Table 1.1-1 2. 1(2)PT-01.7 Heatup/Cooldown Monitoring

3. NEI 99-01 CU3 I OPEP-02.2.1 Rev. 6 1 Page 111 of 303 ATTACHMENT 1 Page 86 of 210 EAL Bases Category: Subcategory:

C -Cold Shutdown / Refueling System Malfunction 3 -RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL: CU3.2 Unusual Event Loss of all RCS temperature and RPV level indication for >- 15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): None BNP Basis: RPV water level is normally monitored using the instruments in Figure C-1 (ref. 1).Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (2121F, ref. 2). These include (ref. 3):* Recirculation Suction Temperatures read on B32-TR-R650 located on panel P-603 (if recirculation loop is in operation)" RHR HX 2A(B) Inlet Temperature as read on E41-TR-R605 Point 1(2), on Panel H12-P614 (RHR HX in service)" RHR HX 2A(B) Outlet Temperature as read on E41-TR-R605 Point 3(4), on Panel H12-P614 (RHR HX not in service)* PPC Display 815, RPV HEATUP/COOLDOWN MONITOR (natural circulation) I OPEP-02.2.1 Rev. 6 1 Page 112 of 303 ATTACHMENT 1 Page 87 of 210 EAL Bases NEI 99-01 Basis: This EAL addresses the inability to determine RCS temperature and RPV level, and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Site Emergency Coordinator should also refer to IC CA3.This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.BNP Basis Reference(s):

1. SD-011.2 Reactor Vessel Instrumentation Figure 01.2-1 Reactor Water Level Instrument Ranges 2. Technical Specifications Table 1.1-1 3. 1(2)PT-01.7 Heatup/Cooldown Monitoring

4. NEI 99-01 CU3 I OPEP-02.2.1 Rev. 6 1 Page 113 of 303 ATTACHMENT 1 Page 88 of 210 EAL Bases Figure C-1 RPV Levels (ref. 1)Reactor Water Level Instrument Ranges 917" 4 FROM VESSEL ZERO850'I OPEP-02.2.1 Rev. 6 1 Page 114 of303 ATTACHMENT 1 Page 89 of 210 EAL Bases Category: Subcategory:

C -Cold Shutdown / Refueling System Malfunction 3 -RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL: CA3.1 Alert UNPLANNED increase in RCS temperature to > 2121F for > Table C-3 duration (Note 1)OR UNPLANNED RPV pressure increase > 10 psig due to a loss of RCS cooling Note 1: The SEC should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Table C-3: RCS Heat-up Duration Thresholds RCS Status Containment Closure Heat-up Duration Status Intact N/A 60 min.*established 20 min.*Not intact Notintactnot established 0 min.* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable. Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): CONTAINMENT CLOSURE -The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to BNP, Containment Closure is established when either Primary Containment is Operable per Section 3.6.1.1 of Technical Specifications or Secondary Containment is considered functional per the requirements of 0AP-022, BNP Outage Risk Management. I OPEP-02.2.1 Rev. 6 1 Page 115 of303 ATTACHMENT 1 Page 90 of 210 EAL Bases 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.BNP Basis: A 10 psig RPV pressure increase can be read on (ref. 1): " Indicator PI-R605A located on Panel P603" Indicator PI-R605B located on Panel P601" Recorder LPR-R608 located on P603* Indicator C32-PI-3332 located on the Remote Shutdown Panel Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (212 0 F, ref. 2). These include (ref. 3): " Recirculation Suction Temperatures read on B32-TR-R650 located on panel P-603 (if recirculation loop is in operation)

RHR HX 2A(B) Inlet Temperature as read on E41-TR-R605 Point 1(2), on Panel H12-P614 (RHR HX in service)* RHR HX 2A(B) Outlet Temperature as read on E41-TR-R605 Point 3(4), on Panel H12-P614 (RHR HX not in service)* PPC Display 815, RPV HEATUP/COOLDOWN MONITOR (natural circulation)

NEI 99-01 Basis: This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE 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.I OPEP-02.2.1 I Rev. 6 Page 116 of 303 ATTACHMENT 1 Page 91 of 210 EAL Bases The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature 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 CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes).This is because 1) the evaporated reactor coolant may be released directly into the Primary Containment or Reactor Building atmosphere and subsequently to the environment, and 2)there is reduced reactor coolant inventory above the top of irradiated fuel.The RCS pressure increase threshold provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability. Escalation of the emergency classification level would be via IC CS1 or RS1.BNP Basis Reference(s):

1. Reactor Vessel Instrumentation System Description SD-01.2 2. BNP Technical Specifications Table 1.1-1 3. 1(2)PT-01.7 Heatup/Cooldown Monitoring

4. Technical Specifications Sections 3.6.1.1 and 3.6.4.1 5. OAP-022, BNP Outage Risk Management

6. NEI 99-01 CA3 I OPEP-02.2.1 Rev. 6 1 Page 117 of 303 ATTACHMENT 1 Page 92 of 210 EAL Bases Category: Subcategory:

C -Cold Shutdown / Refueling System Malfunction 4 -Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL: CU4.1 Unusual Event< 105 VDC bus voltage indications on Technical Specification required 125 VDC buses for> 15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 4 -Cold Shutdown, 5 -Refuel Definition(s): None BNP Basis: There are two independent divisions per unit, designated Division I and Division II (Batteries 1(2)A-1 and 1(2)A-2 for Division I and Batteries 1(2)B-1 and 1(2)B-2 for Division II). Each division consists of a 250 VDC battery center tapped to form two 125 VDC batteries. Each 125 VDC battery has an associated full capacity battery charger. The chargers are supplied from the same AC load groups for which the associated DC subsystem supplies the control power.During normal operation, the DC loads are powered from the battery chargers with the batteries floating on the system. In case of loss of normal power to the battery charger, the DC loads are automatically powered from the station batteries. 105 VDC is the minimum design voltage limit (ref. 1).Note that the Control Room DC voltage indicator only reads battery charger output voltage and not battery voltage unless the charger output breaker is closed. However ERFIS does provide DC battery voltage, otherwise battery voltage must be read locally.This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS7.1.IOPEP-02.2.1 I Rev. 6 1 Page 118 of 303] ATTACHMENT 1 Page 93 of 210 EAL Bases NEI 99-01 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 Division I is out-of-service (inoperable) for scheduled outage maintenance work and Division II is in-service (operable), then a loss of Vital DC power affecting Division II would require the declaration of an Unusual Event. A loss of Vital DC power to Division I would not warrant an emergency classification. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Depending upon the event, escalation of the emergency classification level would be via IC CA1 or CA3, or an IC in Recognition Category R.BNP Basis Reference(s):

1. BNP Technical Specification Bases B.3.8.4 2. OAOP-39.0 LOSS OF DC POWER 3. NEI 99-01 CU4 I OPEP-02.2.1 Rev. 6 1 Page 119 of 303 ATTACHMENT 1 Page 94 of 210 EAL Bases Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 5 -Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: CU5.1 Unusual Event Loss of all Table C-4 onsite communication methods OR Loss of all Table C-4 Offsite communication methods OR Loss of all Table C-4 NRC communication methods Table C-4 Communication Methods System Onsite Offsite NRC Public Address System X PBX Telephone System X X X Corporate Telephone X X X Communications System Commercial Telephones X X X Satellite Phones X X Cellular Phones X X NRC Emergency X Telecommunications System Mode Applicability: 4 -Cold Shutdown, 5 -Refuel, D -Defueled IOPEP-02.2.1 I Rev. 6 Page 120 of 303 ATTACHMENT 1 Page 95 of 210 EAL Bases Definition(s): None BNP Basis: Onsite/offsite/NRC communications include one or more of the systems listed in Table C-4 (ref. 1).Public Address System The Brunswick Plant public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature. This system is powered from the plant uninterruptible power supply which employs battery reserve as well as diesel generator emergency supply.PBX Telephone System The Brunswick Site PBX telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code. The PBX telephone system also provides for outside communications. The PBX switch located in the TSC/EOF building is also backed up by a battery UPS capable of supplying power for a minimum of 8 hours and is augmented by a Diesel Generator capable of supplying power to the TSC/EOF building for at least 5 days.Corporate Telephone Communications System (Voicenet and/or DEMNET)Interconnected through the site PBX and the emergency telephone system, the Voicenet system provides a means to communicate with other corporate locations with which the plant has a need to communicate. This system bypasses external commercial telephone lines and switching equipment. Corporate transmission facilities provide fiber optic, copper-wire, and microwave radio to ensure a high degree of system reliability. In addition to the redundancy provided by the three system options, backup power is provided for the systems. DEMNET is the primary means of offsite communication. This circuit allows intercommunication among the EOF, TSC, control room, counties, and states. DEMNET operates as an internet based (VoIP)communications system with a satellite back-up. Should the internet transfer rate become slow or unavailable, the DEMNET will automatically transfer to satellite mode.IOPEP-02.2.1 Rev. 6 Page 121 of303 ATTACHMENT 1 Page 96 of 210 EAL Bases Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by Duke Energy in four ways: (1) tie-ins through the PBX to any other plant location, (2) lines to plant emergency facilities, (3) lines to the Joint Information Center for public information purposes, and (4) lines to the AEF. The local service provider provides primary and secondary power for their lines at the Central Office.Satellite Phones A total of three portable satellite telephones are available which enable communication when all other phone systems are inoperable, e.g. following a major hurricane. These portable systems can be powered by internal batteries, external DC sources as well as external AC sources. Two of these phones require outside use, while one phone may used either outside or in the EOF with a permanently mounted external antenna.Cellular Phones Selected plant personnel are provided with cellular telephones. These phones may be used during emergencies if other communications means are not readily available or are inoperable. These phones are not expected to be used in the Control Room or Power Block due to interference with plant equipment and loss of signal to the phone.NRC Emergency Telecommunications System The NRC uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRC regional and national offices. The Duke Energy communications line provides a link independent of the local public telephone network.Telephones connected to this network are located in the Brunswick Control Room, Technical Support Center, and Emergency Operations Facility. These lines will not function if the PBX Telephone System fails.This EAL is the cold condition equivalent of the hot condition EAL SU7.1.NEI 99-01 Basis: This IC addresses a significant loss of onsite 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 onsite information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).OPEP-02.2.1 Rev. 6 Page 122 of 303 ATTACHMENT 1 Page 97 of 210 EAL Bases The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations. The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the State, Brunswick and New Hanover County EOCs The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. BNP Basis Reference(s):

1. OERP Radiological Emergency Response Plan Appendix A 2. SD-48 Communication Systems 3. NEI 99-01 CU5 I OPEP-02.2.1 I Rev. 6 Page 123 of 303 ATTACHMENT 1 Page 98 of 210 EAL Bases Category: Subcategory:

Initiating Condition: C -Cold Shutdown / Refueling System Malfunction 6 -Hazardous Event Affecting Safety Systems Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL: CA6.1 Alert The occurrence of any Table C-5 hazardous event AND 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 Table C-5 Hazardous Events* Seismic event (earthquake)

Internal or external FLOODING event* High winds or tornado strike* FIRE* EXPLOSION* Other events with similar hazard characteristics as determined by the Shift Manager Mode Applicability:

4 -Cold Shutdown, 5 -Refuel IOPEP-02.2.1 I Rev. 6 Page 124 of 303 ATTACHMENT 1 Page 99 of 210 EAL Bases Definition(s): 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 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.FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.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 (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. 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. BNP Basis: " The significance of seismic events are discussed under EAL HU2.1 (ref. 1).* Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 2, 3).* Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 135 mph. (ref. 4).I OPEP-02.2.1 Rev. 6 Page 125 of 303 ATTACHMENT 1 Page 100 of 210 EAL Bases* Areas containing functions and systems required for safe shutdown of the plant are identified by Fire Zone in the fire response procedure (ref. 5, 6)." An explosion that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL.NEI 99-01 Basis: This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode.This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.The first conditional 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 conditional 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 CS1 or RS1.BNP Basis Reference(s):

1. 1(2)APP-UA-28 6-4 Seismic Event 2. OAOP-13.0 Operations During Hurricane, Flood Conditions, Tornado or Earthquake

3. Updated FSAR section 3.4.2 Protection From Internal Flooding 4. Updated FSAR Section 2.3.1.2.7 5. BNP-E-9.004 Safe Shutdown Analysis Report 6. OPFP-PBAA Power Block Auxiliary Areas Prefire Plan 7. NEI 99-01 CA6 0PEP-02.2.1 Rev. 6 Page 126 of 303 ATTACHMENT 1 Page 101 of 210 EAL Bases Category H -Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety.1. Security Unauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.3. Natural or Technology Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.

4. Fire Fires can pose significant hazards to personnel and reactor safety. Appropriate for classification are fires within the site Protected Area or which may affect operability of equipment needed for safe shutdown 5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant.IOPEP-02.2.1 I Rev. 6 Page 127of303 ATTACHMENT 1 Page 102 of 210 EAL Bases 6. Control Room Evacuation Events that are indicative of loss of Control Room habitability.

If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.

7. SEC Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification.

While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Site Emergency Coordinator the latitude to classify emergency conditions consistent with the established classification criteria based upon Site Emergency Coordinator judgment.I OPEP-02.2.1 I Rev. 6 Page 128 of 303 ATTACHMENT 1 Page 103 of 210 EAL Bases Category: H -Hazards Subcategory: 1 -Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL: HUI.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervision Mode Applicability: All Definition(s): 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.HOSTILE ACTION -An act toward BNP 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 BNP. 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).BNP Basis: The Security Shift Supervision is defined as either the Security Shift Lieutenant or the Security Shift Sergeant.This EAL is based on the BNP Physical Security Plan (ref. 1).IOPEP-02.2." I Rev. 6 Page 129 of 303 ATTACHMENT 1 Page 104 of 210 EAL Bases NEI 99-01 Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAl, HS1 and HG1.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 Offsite Response Organizations. 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.This EAL references the Security Shift Supervison because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR §2.39 information. 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 BNP Physical Security Plan (ref. 1).Escalation of the emergency classification level would be via IC HA1.BNP Basis Reference(s):

1. BNP Physical Security Plan 2. SEC-NGGC-2170 Security Event Procedures

3. OAOP-40.0 Security Events 4. NEI 99-01 HU1 I OPEP-02.2.1 Rev. 6 1 Page 130 of 303 ATTACHMENT 1 Page 105 of 210 EAL Bases Category: Subcategory:

H -Hazards 1 -Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL: HUl.2 Unusual Event Notification of a credible security threat directed at the site Mode Applicability: All Definition(s): None BNP Basis: The Security Shift Supervision is defined as either the Security Shift Lieutenant or the Security Shift Sergeant.This EAL is based on the BNP Physical Security Plan (ref. 1).NEI 99-01 Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HG1.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 Offsite Response Organizations. 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.I OPEP-02.2.1 Rev. 6 1 Page 131 of303 ATTACHMENT 1 Page 106 of 210 EAL Bases This EAL addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the BNP Physical Security Plan (ref. 1).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 BNP Physical Security Plan (ref. 1).Escalation of the emergency classification level would be via IC HAl.BNP Basis Reference(s):

1. BNP Physical Security Plan 2. SEC-NGGC-2170 Security Event Procedures

3. OAOP-40.0 Security Events 4. NEI 99-01 HU1 I OPEP-02.2.1 Rev. 6 1 Page 132 of 303 ATTACHMENT 1 Page 107 of 210 EAL Bases Category: Subcategory:

H -Hazards 1 -Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL: HUl.3 Unusual Event A validated notification from the NRC providing information of an aircraft threat Mode Applicability: All Definition(s): None BNP Basis: The Security Shift Supervision is defined as either the Security Shift Lieutenant or the Security Shift Sergeant.This EAL is based on the BNP Physical Security Plan (ref. 1).NEI 99-01 Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HG1.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 Offsite Response Organizations. 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.IOPEP-02.2.1 Rev. 6 Page 133 of 303 ATTACHMENT 1 Page 108 of 210 EAL Bases This EAL addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with the BNP Physical Security Plan (ref. 1 ).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 BNP Physical Security Plan (ref. 1).Escalation of the emergency classification level would be via IC HAl.BNP Basis Reference(s):

1. BNP Physical Security Plan 2. SEC-NGGC-2170 Security Event Procedures

3. OAOP-40.0 Security Events 4. NEI 99-01 HU1 OPEP-02.2.1 Rev. 6 Page 134 of 303 ATTACHMENT 1 Page 109 of 210 EAL Bases Category: Subcategory:

Initiating Condition: H -Hazards 1 -Security Hostile action within the owner controlled area or airborne attack threat within 30 minutes EAL: HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervision Mode Applicability: All Definition(s): HOSTILE ACTION -An act toward BNP 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 BNP. 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 -Area depicted as the property boundary in BNP Radiological Emergency Response Plan Figure 1-1.1 Brunswick Site Plan.BNP Basis: The Security Shift Supervision is defined as either the Security Shift Lieutenant or the Security Shift Sergeant.IOPEP-02.2.1 I Rev. 6 1 Page 135 of303 ATTACHMENT 1 Page 110 of 210 EAL Bases NEI 99-01 Basis: This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.Timely and accurate communications between the 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.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 (OROs), allowing them to be better prepared should it be necessary to consider further actions.This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.This EAL is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA.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 BNP Physical Security Plan (ref. 1).BNP Basis Reference(s):

1. BNP Physical Security Plan 2. SEC-NGGC-2170 Security Event Procedures

3. OAOP-40.0 Security Events 4. NEI 99-01 HA1 OPEP-02.2.1 Rev. 6 Page 136 of 303 ATTACHMENT 1 Page 111 of 210 EAL Bases Category: Subcategory:

Initiating Condition: H -Hazards 1 -Security Hostile action within the owner controlled area or airborne attack threat within 30 minutes EAL: HA1.2 Alert A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability: All Definition(s): None BNP Basis: The Security Shift Supervision is defined as either the Security Shift Lieutenant or the Security Shift Sergeant.NEI 99-01 Basis: This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.Timely and accurate communications between the 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.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 (OROs), allowing them to be better prepared should it be necessary to consider further actions.I OPEP-02.2.1 Rev. 6 1 Page 137 of 303 ATTACHMENT 1 Page 112 of 210 EAL Bases This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.This EAL addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with site-specific security 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 BNP Physical Security Plan (ref. 1).BNP Basis Reference(s):

1. BNP Physical Security Plan 2. SEC-NGGC-2170 Security Event Procedures

3. OAOP-40.0 Security Events 4. NEI 99-01 HA1 IOPEP-02.2.1 I Rev. 6 Page 138 of 303 ATTACHMENT 1 Page 113 of 210 EAL Bases Category: Subcategory:

H -Hazards 1 -Security Initiating Condition: Hostile Action within the Protected Area EAL: HSI.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision Mode Applicability: All Definition(s): HOSTILE ACTION -An act toward BNP 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 BNP. 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 double-fenced security area with intrusion detection devices immediately surrounding the plant structures. The Protected Area is depicted in BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations. BNP Basis: The Security Shift Supervision is defined as either the Security Shift Lieutenant or the Security Shift Sergeant.These individuals are the designated onsite personnel qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the BNP Physical Security Plan (Safeguards) information. (ref. 1)IOPEP-02.2.1 I Rev. 6 1 Page 139 of 303 ATTACHMENT 1 Page 114 of 210 EAL Bases NEI 99-01 Basis: This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA.This event will require rapid response and assistance due to the possibility for damage to plant equipment. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan.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 Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions. This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.Emergency plans and implementing procedures are public documents; therefore, EALs 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 BNP Physical Security Plan (ref. 1).Escalation of the emergency classification level would be via IC HG1.BNP Basis Reference(s):

1. BNP Physical Security Plan 2. SEC-NGGC-2170 Security Event Procedures

3. OAOP-40.0 Security Events 4. NEI 99-01 HS1 I 0PEP-02.2.1 Rev. 6 1 Page 140 of 303 ATTACHMENT 1 Page 115 of 210 EAL Bases Category: Subcategory:

H -Hazards 1 -Security Initiating Condition: Hostile Action resulting in loss of physical control of the facility EAL: HGI.1 General Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision AND EITHER of the following has occurred: Any of the following safety functions cannot be controlled or maintained" Reactivity

RPV water level" RCS heat removal OR Damage to spent fuel has occurred or is IMMINENT Mode Applicability:

All Definition(s): HOSTILE ACTION -An act toward BNP 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 BNP. 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).I OPEP-02.2.1 I Rev. 6 Page 14 o43031 ATTACHMENT 1 Page 116 of 210 EAL Bases 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 double-fenced security area with intrusion detection devices immediately surrounding the plant structures. The Protected Area is depicted in BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations. BNP Basis: The Security Shift Supervision is defined as either the Security Shift Lieutenant or the Security Shift Sergeant.NEI 99-01 Basis: This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions. It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan.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 BNP Physical Security Plan (ref.1).BNP Basis Reference(s):

1. BNP Physical Security Plan 2. SEC-NGGC-2170 Security Event Procedures

3. OAOP-40.0 Security Events 4. NEI 99-01 HG1 I OPEP-02.2.1 I Rev. 6 Page 142 of 303 ATTACHMENT 1 Page 117 of 210 EAL Bases Category: Subcategory:

H -Hazards and Other Conditions Affecting Plant Safety 2 -Seismic Event Initiating Condition: Seismic event greater than OBE levels EAL: HU2.1 Unusual Event Seismic event > OBE per OAOP-1 3.0 Mode Applicability: All Definition(s): None BNP Basis: Ground motion acceleration of 0.08g is the Operating Basis Earthquake for BNP (ref. 1).Unit 2 has an active Kinemetrics Condor Seismic Monitoring System with the following components used for seismic detection for the Brunswick Site: The system will detect and digitally record the response to actual earthquake loading in terms of acceleration time history from the existing accelerometers mounted in the Unit 2 -17ft. elevation (basement) of the Reactor Building and also at +89 foot elevation mounted on the Reactor Containment structure. The system will automatically evaluate the recorded acceleration time history in order to determine the response spectra of the events and compare those to the Operating Basis Earthquake (OBE) parameters graphically. It will also determine the exceedance of the OBE, and provides a hard copy of this comparison. The system will provide an immediate Event Alarm output signal at a trigger threshold value of 0.01g to alarm the existing Annunciator 1(2)UA-28 6-4 SEISMIC EVENT in the Control Room back to alert the Operators to a seismic event. (ref. 1, 2)The BNP seismic instrumentation supports readily assessable OBE indications (> 0.08g acceleration) within the Control Room at panel 2-ENV-XU-823. OAOP-13.0 provides the guidance for determining if the OBE earthquake threshold is exceeded. (ref. 3).The Shift Manager or Site Emergency Coordinator 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. OPEP-02.2.1 Rev. 6 Page 143 of 303 ATTACHMENT 1 Page 118 of 210 EAL Bases To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center) can confirm that an earthquake has occurred in the area of the plant. As stated above, such confirmation should not, however, preclude a timely emergency declaration. The NEIC can be contacted by calling (303) 273-8500. Select option #1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of BNP. Provide the analyst with the following BNP coordinates: 330 57' 30" north latitude, 780 00' 30" west longitude (ref. 4). Alternatively, near real-time seismic activity can be accessed via the NEIC website: h ttp ://earthquake. usgs. gov/eqcenter/ NEI 99-01 Basis: This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.Event verification with external sources should not be necessary during or following an OBE.Earthquakes of this magnitude should be readily felt by onsite personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.08g). The Shift Manager or Site Emergency Coordinator 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 SA8.BNP Basis Reference(s):

1. Updated FSAR section 2.5.2.6 2. 1(2)APP-UA-28 6-4 Seismic Event 3. OAOP-13.0 Operations During Hurricane, Flood Conditions, Tornado or Earthquake

4. Updated FSAR section 2.1.1.1 5. NEI 99-01 HU2 I OPEP-02.2.1 Rev. 6 Page 144 of 303 ATTACHMENT 1 Page 119 of 210 EAL Bases Category: Subcategory:

H -Hazards and Other Conditions Affecting Plant Safety 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability: All Definition(s): PROTECTED AREA -The double-fenced security area with intrusion detection devices immediately surrounding the plant structures. The Protected Area is depicted in BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations. BNP Basis: If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or SA8.1.A tornado striking (touching down) within the Protected Area warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm. NEI 99-01 Basis: 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.Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.BNP Basis Reference(s):

1. NEI 99-01 HU3 0PEP-02.2.1 Rev. 6 Page 145 of 303 ATTACHMENT 1 Page 120 of 210 EAL Bases Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.2 Unusual Event 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 Mode Applicability: All Definition(s): FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.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 (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. BNP Basis: Refer to Updated FSAR section 3.4.2 Protection From Internal Flooding to identify susceptible internal flooding areas (ref. 1).IOPEP-02.2.1 I Rev. 6 Page 146 of 303 ATTACHMENT 1 Page 121 of 210 EAL Bases NEI 99-01 Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.This EAL 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.Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.BNP Basis Reference(s):

1. Updated FSAR section 3.4.2 Protection From Internal Flooding 2. NEI 99-01 HU3 I OPEP-02.2.1 I Rev. 6 Page 147 of 303 ATTACHMENT 1 Page 122 of 210 EAL Bases Category: Subcategory:

H -Hazards and Other Conditions Affecting Plant Safety 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.3 Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)Mode Applicability: All Definition(s): IMPEDE(D)- Personnel access to a room or area is hindered to an extent that 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). PROTECTED AREA -The double-fenced security area with intrusion detection devices immediately surrounding the plant structures. The Protected Area is depicted in BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations. BNP Basis: The following documents provide additional information on hazardous substances and spills.* OAOP-34.0 Chlorine Emergencies (Ref. 1)* OAOP-44.0 Sodium Hypochlorite or Acti-Brom Leak Response (Ref. 2)" OAOP-43.0 Hydrogen Emergency (Ref. 3)* OAOP-05.0 Radioactive Spills, High Radiation, and Airborne Activity (Ref. 4)" Regulatory Guide 1.78, Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release, Table 1, Toxicity Limits (IDLH Limits) for Some Hazardous Chemicals (Ref. 5)OPEP-02.2.1 Rev. 6 Page 148 of 303 ATTACHMENT 1 Page 123 of 210 EAL Bases NEI 99-01 Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.This EAL addresses a hazardous materials event originating external to the PROTECTED AREA and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.BNP Basis Reference(s):

1. OAOP-34.0 Chlorine Emergencies

2. OAOP-44.0 Sodium Hypochlorite or Acti-Brom Leak Response 3. OAOP-43.0 Hydrogen Emergency 4. OAOP-05.0 Radioactive Spills, High Radiation, and Airborne Activity 5. Regulatory Guide 1.78, Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release, Table 1, Toxicity Limits (IDLH Limits)for Some Hazardous Chemicals 6. NEI 99-01 HU3 OPEP-02.2.1 Rev. 6 Page 149 of303 ATTACHMENT 1 Page 124 of 210 EAL Bases Category: Subcategory:

H -Hazards and Other Conditions Affecting Plant Safety 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.4 Unusual Event A hazardous event that results in onsite conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7)Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. Mode Applicability: All Definition(s): None BNP Basis: The 15 minute clocks starts when the ORO Director of Emergency Services (Brunswick and New Hanover Counties) and the Shift Manager agree that Onsite/Offsite conditions are sufficient to prohibit the plant staff from accessing the site via personal vehicles.NEI 99-01 Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.I OPEP-02.2.1 I Rev. 6 1 Page 150 of 303 ATTACHMENT 1 Page 125 of 210 EAL Bases This EAL addresses a hazardous event, either onsite or offsite, that causes an onsite 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 onsite 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.Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.BNP Basis Reference(s):

1. NEI 99-01 HU3 IOPEP-02.2.1 I Rev. 6 Page 151 of 303 ATTACHMENT 1 Page 126 of 210 EAL Bases Category: Subcategory:

H -Hazards and Other Conditions Affecting Plant Safety 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.5 Unusual Event Intake Canal water level > +19 ft Mean Sea Level OR Intake Canal water level < -7.75 ft Mean Sea Level Mode Applicability: All BNP Basis: The high Intake Canal level is the highest remotely measurable Intake Canal water level.Otherwise it would have been based based the plant design that Class I structures and engineered safety features systems are protected against still water flooding (elevation 22.0 feet). BNP is geographically located in close proximity to the Atlantic coastal storm track and has an approximate grade elevation of 20 feet above Mean Sea Level. Hurricanes and tropical storms are therefore, the most extreme weather phenomena that affect the site area. Potential subsequent flooding should be considered even though the plant structures were designed to compensate, via installed sump pumps, for a maximum site flooding depth of 22 feet above Mean Sea Level during the Maximum Probable Hurricane. (ref. 1).The minimum water level predicted for the Maximum Probable Hurricane is -7.5 feet Mean Sea Level under special case circumstances. The abnormal operating procedure for a hurricane requires that each unit be shutdown prior to arrival of hurricane conditions at the site. The SW System has been analyzed in modes 4 and 5 for an intake canal water level of -7.75 feet Mean Sea Level corresponding to -8.63 feet Mean Sea Level in the pump suction bay for the maximum pressure drop, 0.88 feet, across the traveling screens. (ref. 2, 3).I OPEP-02.2.1 Rev. 6 1 Page 152 of 303 1 ATTACHMENT 1 Page 127 of 210 EAL Bases NEI 99-01 Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.This EAL addresses high and low external water levels as a result of a hurricane. Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.BNP Basis Reference(s):

1. Updated FSAR section 2.4.10.2 2. Updated FSAR section 9.2.1.2.3 3. OAOP-13.0 Operations During Hurricane, Flood Conditions, Tornado or Earthquake I OPEP-02.2.1 Rev. 6 Page 153 of 303 ATTACHMENT 1 Page 128 of 210 EAL Bases Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 4- Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1): " 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 The FIRE is located within any Table H-1 area Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table H-1 Fire Areas" Reactor Building* Diesel Generator Building" Diesel 4-Day Tank Rooms* Service Water Building* Turbine Building* Control Building* CSTs* Diesel Fuel Oil Storage Tank Mode Applicability:

All IOPEP-02.2.1 I Rev. 6 Page 154 of 303 ATTACHMENT 1 Page 129 of 210 EAL Bases Definition(s): 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.BNP Basis: The 15 minute requirement begins with a credible notification that a fire is occurring, or receipt of multiple valid fire detection system alarms or field validation of a single fire alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 15 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, the 15 minute time limit is from the original receipt of the fire detection alarm.Table H-1 Fire Areas are based on BNP-E-9.004 Safe Shutdown Analysis Report and OPFP-PBAA Power Block Auxiliary Areas Prefire Plan. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS)(ref. 1,2).NEI 99-01 Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.For EAL HU4.1 the intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8.BNP Basis Reference(s):

1. BNP-E-9.004 Safe Shutdown Analysis Report 2. OPFP-PBAA Power Block Auxiliary Areas Prefire Plan 3. NEI 99-01 HU4 OPEP-02.2.1 Rev. 6 Page 155 of 303 ATTACHMENT 1 Page 130 of 210 EAL Bases Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 -Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE)AND The fire alarm is indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table H-1 Fire Areas" Reactor Building" Diesel Generator Building" Diesel 4-Day Tank Rooms" Service Water Building" Turbine Building* Control Building* CSTs" Diesel Fuel Oil Storage Tank Mode Applicability: All OPEP-02.2.1 Rev. 6 Page 156 of 303 ATTACHMENT 1 Page 131 of 210 EAL Bases Definition(s): 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.BNP Basis: The 30 minute requirement begins upon receipt of a single valid fire detection system alarm.The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1.Table H-1 Fire Areas are based on BNP-E-9.004 Safe Shutdown Analysis Report and OPFP-PBAA Power Block Auxiliary Areas Prefire Plan. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS)(ref. 1,2).NEI 99-01 Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.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 HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. IOPEP-02.2.1 I Rev. 6 1 Page 157 of 303 ATTACHMENT 1 Page 132 of 210 EAL Bases Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8.BNP Basis Reference(s):

1. BNP-E-9.004 Safe Shutdown Analysis Report 2. OPFP-PBAA Power Block Auxiliary Areas Prefire Plan 3. NEI 99-01 HU4 IOPEP-02.2.1 I Rev. 6 Page 158 of 303 ATTACHMENT 1 Page 133 of 210 EAL Bases Category: Subcategory:

H -Hazards and Other Conditions Affecting Plant Safety 4 -Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.3 Unusual Event A FIRE within the plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: All Definition(s): 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.PROTECTED AREA -The double-fenced security area with intrusion detection devices immediately surrounding the plant structures. The Protected Area is depicted in BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations. BNP Basis: None NEI 99-01 Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety.I 0PEP-02.2.1 Rev. 6 Page 159 of 303 ATTACHMENT 1 Page 134 of 210 EAL Bases Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8.BNP Basis Reference(s):

1. NEI 99-01 HU4 I OPEP-02.2.1 I Rev. 6 1 Page 160 of 303 ATTACHMENT 1 Page 135 of 210 EAL Bases Category: Subcategory:

H -Hazards and Other Conditions Affecting Plant Safety 4 -Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.4 Unusual Event A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability: All Definition(s): 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.PROTECTED AREA -The double-fenced security area with intrusion detection devices immediately surrounding the plant structures. The Protected Area is depicted in BNP Radiological Emergency Response Plan Figure 1-1.3 Brunswick Site Building and Onsite Emergency Facility Locations. BNP Basis: 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. IOPEP-02.2.1 i Rev. 6 1 Page 161 of 303 ATTACHMENT 1 Page 136 of 210 EAL Bases NEI 99-01 Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.If a FIRE within the plant PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8.BNP Basis Reference(s):

1. NEI 99-01 HU4 IOPEP-02.2.1 I Rev. 6 Page 162 of 303 ATTACHMENT 1 Page 137 of 210 EAL Bases Category: Subcategory:

H -Hazards and Other Conditions Affecting Plant Safety 5 -Hazardous Gases Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL: HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 rooms or areas AND Entry into the room or area is prohibited or IMPEDED (Note 5)Note 5: 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. Table H-2 Safe Operation & Shutdown Areas Room/Area Mode Applicability Reactor Building -17' North RHR Unit-1 & 2 3, 4, 5 Reactor Building -17' South RHR Unit-1 & 2 3, 4, 5 Reactor Building 20' East & West MCC Areas Unit-1 & 2 3, 4, 5 Reactor Building 20' Pipe Tunnel Unit-1 & 2 3, 4, 5 Mode Applicability: All Definition(s): IMPEDE(D)- Personnel access to a room or area is hindered to an extent that 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). BNP Basis: If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.OPEP-02.2.1 I Rev. 6 1 Page 163 of 303 ATTACHMENT 1 Page 138 of 210 EAL Bases The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).The following documents provide additional information on hazardous substances and spills.* OAOP-34.0 Chlorine Emergencies (Ref. 2)" OAOP-44.0 Sodium Hypochlorite or Acti-Brom Leak Response (Ref. 3)* OAOP-43.0 Hydrogen Emergency (Ref. 4)NEI 99-01 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 Site Emergency Coordinator'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). IOPEP-02.2.1 I Rev. 6 Page 164 of 303 ATTACHMENT 1 Page 139 of 210 EAL Bases An emergency declaration is not warranted if any of the following conditions apply: " The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release). For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4." The gas release is a 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." If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.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 ICs.BNP Basis Reference(s):

1. Attachment 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases 2. OAOP-34.0 Chlorine Emergencies

3. OAOP-44.0 Sodium Hypochlorite or Acti-Brom Leak Response 4. OAOP-43.0 Hydrogen Emergency 5. NEI 99-01 HA5 OPEP-02.2.1 Rev. 6 Page 165 of 303 ATTACHMENT 1 Page 140 of 210 EAL Bases Category: Subcategory:

H -Hazards and Other Conditions Affecting Plant Safety 6 -Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL: HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panels Mode Applicability: All Definition(s): None BNP Basis: The Shift Manager (SM) determines if the Control Room is inoperable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions (Ref. 1, 2).Inability to establish plant control from outside the Control Room escalates this event to a Site Area Emergency per EAL HS6.1.NEI 99-01 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. IOPEP-02.2.1 I Rev. 6 Page 166 of 303] ATTACHMENT 1 Page 141 of 210 EAL Bases Escalation of the emergency classification level would be via IC HS6.BNP Basis Reference(s):

1. 0AOP-32.0, Plant Shutdown from Outside Control Room 2. OPLP-01.5 Alternate Shutdown Capability Controls 3. NEI 99-01 HA6 I OPEP-02.2.1 I Rev. 6 Page 167 of 3031 ATTACHMENT 1 Page 142 of 210 EAL Bases Category: Subcategory:

H -Hazards and Other Conditions Affecting Plant Safety 6 -Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL: HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panels AND Control of any of the following key safety functions is not reestablished within 22.5 min.(Note 1):* Reactivity" RPV water level" RPV pressure Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: All Definition(s): None BNP Basis: The Shift Manager determines if the Control Room is inoperable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions (Ref. 1, 2).The immediate actions of OAOP-32 direct a reactor scram prior to evacuating the Control Room thus no further action is required for reactivity control. Local control of high pressure injection sources and Safety Relief Valves (SRVs) establishes control of RPV water level and pressure.I OPEP-02.2.1 Rev. 6 Page 168 of 303 ATTACHMENT 1 Page 143 of 210 EAL Bases NEI 99-01 Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Site Emergency Coordinator judgment. The Site Emergency Coordinator is expected to make a reasonable, informed judgment within 22.5 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s) (ref. 3).Escalation of the emergency classification level would be via IC FG1 or CG1 BNP Basis Reference(s):

1. OAOP-32.0, Plant Shutdown from Outside Control Room 2. OPLP-01.5 Alternate Shutdown Capability Controls 3. Calculation No. BNP-E-9.007 ASSD Manual Action Feasibility

4. NEI 99-01 HS6 I OPEP-02.2.1 I Rev. 6 Page 169 of 303 ATTACHMENT 1 Page 144 of 210 EAL Bases Category: Subcategory:

H -Hazards and Other Conditions Affecting Plant Safety 7 -SEC Judgment Initiating Condition: Other conditions existing that in the judgment of the Site Emergency Coordinator warrant declaration of a UE EAL: HU7.1 Unusual Event Other conditions exist which in the judgment of the Site Emergency Coordinator 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.Mode Applicability: All Definition(s): None BNP Basis: The Site Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the BNP Emergency Response Plan. The Operations Shift Manager (SM) initially acts in the capacity of the Site Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Site Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref.1).NEI 99-01 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 Site Emergency Coordinator to fall under the emergency classification level description for an Unusual Event.OPEP-02.2.1 Rev. 6 Page 170 of 303 ATTACHMENT 1 Page 145 of 210 EAL Bases BNP Basis Reference(s):

1. OERP BNP Radiological Emergency Response Plan section 3.0 Emergency Response Organization

2. NEI 99-01 HU7 I OPEP-02.2.1 Rev. 6 1 Page 171 of 303 ATTACHMENT 1 Page 146 of 210 EAL Bases Category: Subcategory:

H -Hazards and Other Conditions Affecting Plant Safety 7 -SEC Judgment Initiating Condition: Other conditions exist that in the judgment of the Site Emergency Coordinator warrant declaration of an Alert EAL: HA7.1 Alert Other conditions exist which, in the judgment of the Site Emergency Coordinator, 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.Mode Applicability: All Definition(s): HOSTILE ACTION -An act toward BNP 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 BNP. 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).IOPEP-02.2.1 I Rev. 6 Page 172 of 303 ATTACHMENT 1 Page 147 of 210 EAL Bases BNP Basis: The Site Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the BNP Emergency Response Plan. The Operations Shift Manager (SM) initially acts in the capacity of the Site Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Site Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1 ).NEI 99-01 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 Site Emergency Coordinator to fall under the emergency classification level description for an Alert.BNP Basis Reference(s):

1. OERP BNP Radiological Emergency Response Plan section 3.0 Emergency Response Organization

2. NEI 99-01 HA7 I 0PEP-02.2.1 I Rev. 6 Page 173 of 303 ATTACHMENT 1 Page 148 of 210 EAL Bases Category: Subcategory:

Initiating Condition: EAL: H -Hazards and Other Conditions Affecting Plant Safety 7 -SEC Judgment Other conditions existing that in the judgment of the Site Emergency Coordinator warrant declaration of a Site Area Emergency HS7.1 Site Area Emergency Other conditions exist which in the judgment of the Site Emergency Coordinator 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 Mode Applicability: All Definition(s): HOSTILE ACTION -An act toward BNP 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 BNP. 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).IOPEP-02.2.1 I Rev. 6 1 Page 174 of 303 ATTACHMENT 1 Page 149 of 210 EAL Bases BNP Basis: The Site Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the BNP Emergency Response Plan. The Operations Shift Manager (SM) initially acts in the capacity of the Site Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Site Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref.1).NEI 99-01 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 Site Emergency Coordinator to fall under the emergency classification level description for a Site Area Emergency. BNP Basis Reference(s):

1. OERP BNP Radiological Emergency Response Plan section 3.0 Emergency Response Organization

2. NEI 99-01 HS7 I OPEP-02.2.1 Rev. 6 Page 175 of 303 ATTACHMENT 1 Page 150 of 210 EAL Bases Category: Subcategory:

H -Hazards and Other Conditions Affecting Plant Safety 7 -SEC Judgment Initiating Condition: Other conditions exist which in the judgment of the Site Emergency Coordinator warrant declaration of a General Emergency EAL: HG7.1 General Emergency Other conditions exist which in the judgment of the Site Emergency Coordinator 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 Mode Applicability: All Definition(s): HOSTILE ACTION -An act toward BNP 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 BNP. 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.IOPEP-02.2.1 I Rev. 6 Page 176 of 303 ATTACHMENT 1 Page 151 of 210 EAL Bases BNP Basis: The Site Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the BNP Emergency Response Plan. The Operations Shift Manager(SM) initially acts in the capacity of the Site Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Site Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref.1).Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the Site Boundary.NEI 99-01 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 Site Emergency Coordinator to fall under the emergency classification level description for a General Emergency. BNP Basis Reference(s):

1. OERP BNP Radiological Emergency Response Plan section 3.0 Emergency Response Organization

2. NEI 99-01 HG7 I OPEP-02.2.1 Rev. 6 1 Page 177 of 303 ATTACHMENT 1 Page 152 of 210 EAL Bases Category S -System Malfunction EAL Group: Hot Conditions (RCS temperature

> 212 0 F); EALs in this category are applicable only in one or more hot operating modes.Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety.The events of this category pertain to the following subcategories:

1. Loss of Emergency AC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

This category includes loss of onsite and offsite sources for 4160 V emergency buses.2. Loss of Vital DC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of vital plant 125 VDC power sources.3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.

4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (2% -5% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category.

However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.I OPEP-02.2.1 Rev. 6 1 Page 178 of303 ATTACHMENT 1 Page 153 of 210 EAL Bases 5. RCS Leakage The reactor pressure vessel provides a volume for the coolant that covers the reactor core.The reactor pressure vessel and associated pressure piping (reactor coolant system)together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and Primary Containment integrity.

6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS)to initiate and complete reactor scrams. In the plant licensing basis, postulated failures of the RPS to complete a reactor scram comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any trip failure event that does not achieve reactor shutdown.

If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and Primary Containment integrity.

7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

8. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant safety system performance or significant visible damage warrant emergency classification under this subcategory.

I OPEP-02.2.1 I Rev. 6 Page 179 of 303 ATTACHMENT 1 Page 154 of 210 EAL Bases Category: Subcategory: Initiating Condition: S -System Malfunction 1 -Loss of Emergency AC Power Loss of all offsite AC power capability to emergency buses for 15 minutes or longer EAL: SUI.1 Unusual Event Loss of all offsite AC power capability to Emergency 4 KV Buses El (E3) and E2(E4) for-15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None Basis: BNP Basis: The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. 4160 VAC power is divided into Balance of Plant (BOP Bus), and Emergency (E-Bus) distribution. The BOP Buses consist of the Conventional Buses (Common A/B Buses;Buses 1 B/2B, 1 C/2C, 1 D/2D) and Caswell Beach Buses A/B. The Emergency switchgear are Buses EI/E3 (Division I) and E2/E4 (Division II).The E-Buses are normally powered through the respective BOP Buses (1 D to El, 1 C to E2, 2D to E3, 2C to E4) via a master/slave breaker arrangement. Each E-Bus has a dedicated Diesel Generator to supply an emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of offsite power. The DGs will automatically start and tie onto the E-Buses if the normal power source or offsite power is lost. The DGs can provide power to the E-Buses only. In the event the diesel generator is unavailable for an E-Bus, crosstie capability exists for each E-Bus from the same division of the opposite unit (El to E3, E2 to E4). Although the E-Buses within the unit also have crosstie capability, this alignment is not permitted by plant procedures, with the exception of El to E2 during specific Alternate Safe Shutdown (ASSD) conditions. OPEP-02.2.1 Rev. 6 Page 180 of 303 ATTACHMENT 1 Page 155 of 210 EAL Bases During periods of unit shutdown, when the Startup Auxiliary Transformer (SAT) would be the only normal source of offsite power, the Unit Auxiliary Transformer (UAT) can be made available by establishing a UAT backfeed. Backfeed from the UAT will require the use of keys for the control selector switches and opening of the respective generator's manual no-load disconnect. (Ref. 1, 2)The 15-minute interval was selected as a threshold to exclude transient or momentary power losses.NEI 99-01 Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency buses. This condition represents a potential reduction in the level of safety of the plant.For emergency classification purposes, "capability" means that an offsite AC power source(s)is available to the emergency buses, whether or not the buses are powered from it.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.Escalation of the emergency classification level would be via IC SAI.BNP Basis Reference(s):

1. BNP Updated FSAR Chapter 8 2. 1(2)OP-50 Plant Electric System Operating Procedure 3. NEI 99-01 SUl I OPEP-02.2.1 Rev. 6 1 Page 181 of 303 ATTACHMENT 1 Page 156 of 210 EAL Bases Category:

S -System Malfunction Subcategory: 1 -Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer EAL: SAI.1 Alert AC power capability to Emergency 4 KV Buses El (E3) and E2(E4) reduced to a single power source for -15 min. (Note 1)AND Any additional single power source failure will result in loss of all unit-specific AC power to SAFETY SYSTEMS Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): 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 (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. I OPEP-02.2.1 Rev. 6 Page 182 of 303 ATTACHMENT 1 Page 157 of 210 EAL Bases Basis: BNP Basis: The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. 4160 VAC power is divided into Balance Of Plant (BOP Bus), and Emergency (E-Bus) distribution. The BOP Buses consist of the Conventional Buses (Common A/B Buses;Buses 1 B/2B, 1C/2C, 1 D/2D) and Caswell Beach Buses A/B. The Emergency switchgear are Buses El/E3 (Division I) and E2/E4 (Division II).The E-Buses are normally powered through the respective BOP Buses (1D to El, 1C to E2, 2D to E3, 2C to E4) via a master/slave breaker arrangement. Each E-Bus has a dedicated Diesel Generator to supply an emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of offsite power. The DGs will automatically start and tie onto the E-Buses if the normal power source or offsite power is lost. The DGs can provide power to the E-Buses only. In the event the diesel generator is unavailable for an E-Bus, crosstie capability exists for each E-Bus from the same division of the opposite unit (El to E3, E2 to E4). Although the E-Buses within the unit also have crosstie capability, this alignment is not permitted by plant procedures, with the exception of El to E2 during specific Alternate Safe Shutdown (ASSD) conditions. During periods of unit shutdown, when the Startup Auxiliary Transformer (SAT) would be the only normal source of offsite power, the Unit Auxiliary Transformer (UAT) can be made available by establishing a UAT backfeed. Backfeed from the UAT will require the use of keys for the control selector switches and opening of the respective generator's manual no-load disconnect (Ref. 1, 2).The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. If the capability of a second source of emergency bus power is not restored within 15 minutes, an Alert is declared under this EAL.Because 2 RHR pumps on each unit are powered from the unaffected unit, the words "unit-specific" have been added to clarify that the cross-connected RHR pump power cannot be credited as an AC power source relative to this EAL.NEI 99-01 Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU1.IOPEP-02.2.1 Rev. 6 Page 183of303 ATTACHMENT 1 Page 158 of 210 EAL Bases An "AC power source" is a source recognized in AOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below." A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator)." A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being 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 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 SS1.BNP Basis Reference(s):

1. BNP Updated FSAR Chapter 8 2. 1(2)OP-50 Plant Electric System Operating Procedure 3. NEI 99-01 SA1 I OPEP-02.2.1 Rev. 6 Page 184 of 303 ATTACHMENT 1 Page 159 of 210 EAL Bases Category: Subcategory:

Initiating Condition: S -System Malfunction 1 -Loss of Emergency AC Power Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer EAL: SS1.1 Site Area Emergency Loss of all offsite and all onsite AC power capability to Emergency 4 KV Buses El (E3)and E2(E4) for -15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: This EAL is indicated by the loss of all offsite and onsite AC power capability to Emergency 4 KV Buses El (E3) and E2(E4) for greater than or equal to 15 minutes.The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. 4160 VAC power is divided into Balance Of Plant (BOP Bus), and Emergency (E-Bus) distribution. The BOP Buses consist of the Conventional Buses (Common A/B Buses;Buses 1 B/2B, 1 C/2C, 1 D/2D) and Caswell Beach Buses A/B. The Emergency switchgear are Buses El/E3 (Division I) and E2/E4 (Division II).The E-Buses are normally powered through the respective BOP Buses (1D to El, 1C to E2, 2D to E3, 2C to E4) via a master/slave breaker arrangement. Each E-Bus has a dedicated Diesel Generator to supply an emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of offsite power. The DGs will automatically start and tie onto the E-Buses if the normal power source or offsite power is lost. The DGs can provide power to the E-Buses only. In the event the diesel generator is unavailable for an E-Bus, crosstie capability exists for each E-Bus from the same division of the opposite unit (El to E3, E2 to E4). Although the E-Buses within the unit also have crosstie capability, this alignment is not permitted by plant procedures, with the exception of El to E2 during specific Alternate Safe Shutdown (ASSD) conditions. OPEP-02.2.1 Rev. 6 Page 185 of 303 ATTACHMENT 1 Page 160 of 210 EAL Bases During periods of unit shutdown, when the Startup Auxiliary Transformer (SAT) would be the only normal source of offsite power, the Unit Auxiliary Transformer (UAT) can be made available by establishing a UAT backfeed. Backfeed from the UAT will require the use of keys for the control selector switches and opening of the respective generator's manual no-load disconnect (Ref. 1, 2)The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. The interval begins when both offsite and onsite AC power are lost.NEI 99-01 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 RG1, FG1 or SG1.BNP Basis Reference(s):

1. BNP Updated FSAR Chapter 8 2.3.1(2)OP-50 Plant Electric System Operating Procedure OAOP-36.2 Station Blackout 4. NEI 99-01 SS1 I 0PEP-02.2.1 Rev. 6 1 Page 186 of 303 ATTACHMENT 1 Page 161 of 210 EAL Bases Category: Subcategory:

S -System Malfunction 1 -Loss of Emergency AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses OR loss of all emergency AC and vital DC power sources for 15 minutes or longer EAL: SG1.1 General Emergency Loss of all offsite and all onsite AC power capability to Emergency 4 KV Buses El (E3)and E2(E4)AND EITHER: " Restoration of at least one emergency bus in < 4 hours is not likely (Note 1)" RPV water level cannot be restored and maintained > MSCRWL (LL-4)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: This EAL is indicated by the extended loss of all offsite and onsite AC power capability to 4 KV emergency buses El (E3) and E2(E4) either for greater then the BNP Station Blackout (SBO)coping analysis time (4 hrs.) (ref. 1) or that has resulted in indications of an actual loss of adequate core cooling.Indication of continuing core cooling degradation is manifested by the inability to restore and maintain RPV water level above the Minimum Steam Cooling Reactor Water Level (LL-4) (ref.4, 5).I OPEP-02.2.1 Rev. 6 1 Page 187 of 303 ATTACHMENT 1 Page 162 of 210 EAL Bases The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. 4160 VAC power is divided into Balance of Plant (BOP Bus), and Emergency (E-Bus) distribution. The BOP Buses consist of the Conventional Buses (Common A/B Buses;Buses 1 B/2B, 1 C/2C, 1 D/2D) and Caswell Beach Buses A/B. The Emergency switchgear are Buses El/E3 (Division I) and E2/E4 (Division II).The E-Buses are normally powered through the respective BOP Buses (1 D to El, 1C to E2, 2D to E3, 2C to E4) via a master/slave breaker arrangement. Each E-Bus has a dedicated Diesel Generator to supply an emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of offsite power. The DGs will automatically start and tie onto the E-Buses if the normal power source or offsite power is lost. The DGs can provide power to the E-Buses only. In the event the diesel generator is unavailable for an E-Bus, crosstie capability exists for each E-Bus from the same division of the opposite unit (El to E3, E2 to E4). Although the E-Buses within the unit also have crosstie capability, this alignment is not permitted by plant procedures, with the exception of El to E2 during specific Alternate Safe Shutdown (ASSD) conditions. During periods of unit shutdown, when the Startup Auxiliary Transformer (SAT) would be the only normal source of offsite power, the Unit Auxiliary Transformer (UAT) can be made available by establishing a UAT backfeed. Backfeed from the UAT will require the use of keys for the control selector switches and opening of the respective generator's manual no-load disconnect (Ref. 2, 3).Four hours is the station blackout coping time (ref 1).Indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Site Emergency Coordinator judgment as it relates to imminent Loss or Potential Loss of fission product barriers and degraded ability to monitor fission product barriers. Indication of continuing core cooling degradation is manifested by an RPV level that cannot be restored and maintained > MSCRWL (LL-4) (ref. 4, 5). Core submergence is the most desirable means of core cooling, however when RPV level is below TAF, the uncovered portion of the core can be cooled by less reliable means (i.e., steam cooling or spray cooling).NEI-9901 Basis: This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. I0PEP-02.2.1 I Rev. 6 1 Page 188 of 303 ATTACHMENT 1 Page 163 of 210 EAL Bases The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions.Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of 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 bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.BNP Basis Reference(s):

1. 0AOP-36.2 STATION BLACKOUT, Section 4.0 2. BNP Updated FSAR Chapter 8 3. 1(2)OP-50 Plant Electric System Operating Procedure 4. 1(2)EOP-01 Reactor Vessel Control Procedure 5. OEOP-01-NL, EOP SAMG NUMERICAL LIMITS AND VALUES I OPEP-02.2.1 Rev. 6 Page 189 of 303 ATTACHMENT 1 Page 164 of 210 EAL Bases Category: Subcategory:

Initiating Condition: S -System Malfunction 1 -Loss of Emergency AC Power Prolonged loss of all offsite and all onsite AC power to emergency buses OR loss of all emergency AC and vital DC power sources for 15 minutes or longer EAL: SG1.2 General Emergency Loss of all offsite and all onsite AC power capability to Emergency 4 KV Buses El (E3)and E2(E4) for -> 15 min.AND Loss of all 125 VDC power based on battery bus voltage indications < 105 VDC on all vital DC buses 1(2)A-1, A-2, B-1 and B-2 for >- 15 min.(Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: This EAL is indicated by the loss of all offsite and onsite emergency AC power capability to 4 KV emergency buses El (E3) and E2(E4) for greater than 15 minutes in combination with degraded vital DC power voltage. This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi.The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. 4160 VAC power is divided into Balance of Plant (BOP Bus), and Emergency (E-Bus) distribution. The BOP Buses consist of the Conventional Buses (Common A/B Buses;Buses 1 B/2B, 1 C/2C, 1 D/2D) and Caswell Beach Buses A/B. The Emergency switchgear are Buses EI/E3 (Division I) and E2/E4 (Division II).OPEP-02.2.1 Rev. 6 Page 190 of 303 ATTACHMENT 1 Page 165 of 210 EAL Bases The E-Buses are normally powered through the respective BOP Buses (1D to El, 1C to E2, 2D to E3, 2C to E4) via a master/slave breaker arrangement. Each E-Bus has a dedicated Diesel Generator to supply an emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of offsite power. The DGs will automatically start and tie onto the E-Buses if the normal power source or offsite power is lost. The DGs can provide power to the E-Buses only. In the event the diesel generator is unavailable for an E-Bus, crosstie capability exists for each E-Bus from the same division of the opposite unit (El to E3, E2 to E4). Although the E-Buses within the unit also have crosstie capability, this alignment is not permitted by plant procedures, with the exception of El to E2 during specific Alternate Safe Shutdown (ASSD) conditions. During periods of unit shutdown, when the Startup Auxiliary Transformer (SAT) would be the only normal source of offsite power, the Unit Auxiliary Transformer (UAT) can be made available by establishing a UAT backfeed. Backfeed from the UAT will require the use of keys for the control selector switches and opening of the respective generator's manual no-load disconnect (Ref. 2, 3).There are two independent vital 125 VDC divisions per unit, designated Division I and Division II (Batteries 1(2)A-1 and 1(2)A-2 for Division I and Batteries 1(2)B-1 and 1(2)B-2 for Division II). Each division consists of a 250 VDC battery center tapped to form two 125 VDC batteries. Each 125 VDC battery has an associated full capacity battery charger. The chargers are supplied from the same AC load groups for which the associated DC subsystem supplies the control power. During normal operation, the DC loads are powered from the battery chargers with the batteries floating on the system. In case of loss of normal power to the battery charger, the DC loads are automatically powered from the station batteries. 105 VDC is the minimum design voltage limit (ref. 4, 5).NEI-9901 Basis: This IC addresses a concurrent and prolonged loss of both emergency AC and Vital DC power. A loss of all emergency 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 emergency AC and vital DC power will lead to multiple challenges to fission product barriers.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.I0PEP-02.2.1 I Rev. 6 Page 191 of 303 ATTACHMENT 1 Page 166 of 210 EAL Bases BNP Basis Reference(s):

1. OAOP-36.2 STATION BLACKOUT, Section 4.0 2. BNP Updated FSAR Chapter 8 3. 1(2)OP-50 Plant Electric System Operating Procedure 4. BNP Technical Specification Bases B.3.8.4 5. OAOP-39.0 LOSS OF DC POWER IOPEP-02.2.1 I Rev. 6 1 Page 192 of 303 ATTACHMENT 1 Page 167 of 210 EAL Bases Category: Subcategory:

S -System Malfunction 2 -Loss of Vital DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer EAL: SS2.1 Site Area Emergency Loss of all 125 VDC power based on battery bus voltage indications < 105 VDC on all vital DC buses 1(2)A-1, A-2, B-1 and B-2 for -- 15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: There are two independent vital 125 VDC divisions per unit, designated Division I and Division II (Batteries 1(2)A-1 and 1(2)A-2 for Division I and Batteries 1(2)B-1 and 1(2)B-2 for Division II). Each division consists of a 250 VDC battery center tapped to form two 125 VDC batteries. Each 125 VDC battery has an associated full capacity battery charger. The chargers are supplied from the same AC load groups for which the associated DC subsystem supplies the control power. During normal operation, the DC loads are powered from the battery chargers with the batteries floating on the system. In case of loss of normal power to the battery charger, the DC loads are automatically powered from the station batteries. Note that the Control Room DC voltage indicator only reads battery charger output voltage and not battery voltage unless the charger output breaker is closed. However ERFIS does provide DC battery voltage, otherwise battery voltage must be read locally.105 VDC is the minimum design voltage limit (ref. 1).I OPEP-02.2.1 Rev. 6 1 Page 193 of 303 ATTACHMENT 1 Page 168 of 210 EAL Bases NEI 99-01 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.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level would be via ICs RG1, FG1 or SG1.BNP Basis Reference(s):

1. BNP Technical Specification Bases B.3.8.4 2. OAOP-39.0 LOSS OF DC POWER 3. NEI 99-01 SS8 I OPEP-02.2.1 I Rev. 6 Page 194 of 303 ATTACHMENT 1 Page 169 of 210 EAL Bases Category: Subcategory:

S -System Malfunction 3 -Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL: SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for > 15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table S-1 Safety System Parameters

Reactor power" RPV water level* RPV pressure* Primary containment pressure* Torus water level" Torus temperature Mode Applicability:

1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): 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.BNP Basis: SAFETY SYSTEM parameters listed in Table S-1 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.The ERFIS, which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1, 2).As used in this EAL "within the Control Room" means any available indicator available within the Control Room boundary, including back panels.OPEP-02.2.1 Rev. 6 Page 195 of 303 ATTACHMENT 1 Page 170 of 210 EAL Bases NEI 99-01 Basis: This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital or 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-1 022) 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 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 IC SA3.BNP Basis Reference(s):

1. Updated FSAR Update Section 7.7.1.9 2. 001-01.08 Control of Equipment and System Status 3. NEI 99-01 SU2 I OPEP-02.2.1 I Rev. 6 Page 196 of 303 ATTACHMENT 1 Page 171 of 210 EAL Bases Category: Subcategory:

Initiating Condition: S -System Malfunction 3 -Loss of Control Room Indications UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL: SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for -15 min. (Note 1)AND Any significant transient is in progress, Table S-2 Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table S-1 Safety System Parameters" Reactor power" RPV water level* RPV pressure* Primary containment pressure* Torus water level* Torus temperature Table S-2 Significant Transients

Reactor scram* Runback > 25% rated thermal power* Electrical load rejection

> 25%electrical load* ECCS injection* Thermal power oscillations > 10%(peak to peak)LOPEP-02.2.1 I Rev. 6 1 Page 197 of 303 ATTACHMENT 1 Page 172 of 210 EAL Bases Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): 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.BNP Basis: SAFETY SYSTEM parameters listed in Table S-1 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.The ERFIS, which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1, 2).Significant transients are listed in Table S-2 and include response to automatic or manually initiated functions such as scrams, runbacks (Recirculation) involving greater than 25%thermal power change, electrical load rejections of greater than 25% full electrical load, ECCS injections, or thermal power oscillations of 10% (peak to peak) or greater.As used in this EAL "within the Control Room" means any available indicator available within the Control Room boundary, including back panels.NEI 99-01 Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital or 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-1 022) 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. OPEP-02.2.1 Rev. 6 Page 198 of 303 ATTACHMENT 1 Page 173 of 210 EAL Bases 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 FS1 or IC RSI.BNP Basis Reference(s):

1. Updated FSAR Update Section 7.7.1.9 2. 001-01.08 Control of Equipment and System Status 3. NEI 99-01 SA2 I 0PEP-02.2.1 Rev. 6 1 Page 199 of 303 ATTACHMENT 1 Page 174 of 210 EAL Bases Category:

S -System Malfunction Subcategory: 4 -RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL: SU4.1 Unusual Event Steam Jet Air Ejector Radiation Monitor 1(2)D12-RM-K601A /B Hi-Hi alarm (Process Off-Gas Rad Hi-Hi alarm 1(2)UA-03 4-2) > 15 min. (Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: The Steam Jet Air Ejector radiation monitor setpoint provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 1 OCFR1 00 in the event of an inadvertent release via the condenser air ejector (ref. 2, 3).At the Hi-Hi alarm setpoint, the process Off-Gas timer is started. After the process Off-Gas timer has timed out (15 minutes), the Off-Gas system will isolate (ref. 1).NEI 99-01 Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category R ICs.IOPEP-02.2.1 I Rev. 6 Page 200 of 303 ATTACHMENT 1 Page 175 of 210 EAL Bases BNP Basis Reference(s):

1. ARP 1(2)APP-UA-03 4-2 Process Off-Gas Rad Hi-Hi 2. BNP Offsite Dose Calculation Manual section 3.1.3 3. BNP Technical Specifications section 3.7.5 4. NEI 99-01 SU3 OPEP-02.2.1 Rev. 6 Page 201 of 303 ATTACHMENT 1 Page 176 of 210 EAL Bases Category: Subcategory:

Initiating Condition: S -System Malfunction 4 -RCS Activity Reactor coolant activity greater than Technical Specification allowable limits EAL: SU4.2 Unusual Event Coolant activity > 0.2 pCi/gm 1-131 dose equivalent for > 48 hours OR Coolant activity > 4.0 pCi/gm 1-131 dose equivalent instantaneous Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: The specific iodine activity is limited to -0.2 pCi/gm Dose Equivalent 1-131. This limit ensures the source term assumed in the safety analysis for the Main Steam Line Break (MSLB) is not exceeded, so any release of radioactivity to the environment during an MSLB is less than a small fraction of the 10 CFR 50.67 limits (ref. 1).The upper limit of 4.0 pCi/gm Dose Equivalent 1-131 ensures that the thyroid dose from an MSLB will not exceed the dose guidelines of 10 CFR 50.67 or Control Room operator dose limits specified in GDC 19 of 10 CFR 50, Appendix A (ref. 1).NEI 99-01 Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category R ICs.0PEP-02.2.1 Rev. 6 1 Page 202 of 303 ATTACHMENT 1 Page 177 of 210 EAL Bases BNP Basis Reference(s):

1. BNP Technical Specifications section 3.4.6 2. NEI 99-01 SU3 I OPEP-02.2.1 Rev. 6 1 Page 203 of 303 ATTACHMENT 1 Page 178 of 210 EAL Bases Category:

S -System Malfunction Subcategory: 5- RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL: SU5.1 Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for > 15 min.OR RCS identified leakage > 25 gpm for > 15 min.OR Leakage from the RCS to a location outside Primary Containment > 25 gpm for > 15 min.(Note 1)Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: Leakage is monitored by utilizing the following techniques:

Sensing excess flow in piping systems* Sensing pressure and temperature changes in the primary containment" Monitoring for high flow and temperature through selected drains,* Sampling airborne particulate and gaseous radioactivity.
Drywell floor and equipment drain sump leak rate system I OPEP-02.2.1 Rev. 6 Page 204 of 303 ATTACHMENT 1 Page 179 of 210 EAL Bases Identified leakage is leakage into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or leakage into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary leakage. (ref.1,2)Unidentified leakage is all leakage into the drywell that is not identified leakage. (ref. 1, 2)Pressure boundary leakage is leakage through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall. (ref. 1, 2)The drywell floor drain sump flow monitoring system monitors the leakage collected in the floor drain sump. This unidentified leakage consists of leakage from control rod drives, valve flanges, floor drains, the Reactor Building Closed Cooling Water System, and drywell cooler drains, and any leakage not collected in the drywell equipment drain sump. The drywell floor drain sump is provided with two sump pumps. A flow transmitter in the common discharge line of the drywell floor drain sump pumps inputs to a flow integrator.

In addition to the required instrumentation, the starting frequency and run duration of a sump pump motor are monitored by timer circuitry to provide a signal (alarm) in the Control Room indicating that leakage has reached a specified limit. (ref. 2)RCS leakage outside of the Primary Containment that is not considered identified or unidentified leakage per Technical Specifications includes leakage via interfacing systems such as RCS to the Reactor Building Closed Cooling Water (RBCCW system), or systems that directly see RCS pressure outside primary containment such as Reactor Water Cleanup (RWCU), reactor water sampling system and Residual Heat Removal (RHR) system (when in the shutdown cooling mode) (ref. 3)Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation, EAL FA1.1. The note has been added to remind the EAL-user to review Table F-1 for possible escalation to higher emergency classifications. NEI 99-01 Basis: This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.I OPEP-02.2.1 Rev. 6 Page 205 of 303 ATTACHMENT 1 Page 180 of 210 EAL Bases The first and second EAL conditions 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). The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the Primary Containment, or a location outside of Primary Containment. The leak rate values for each condition 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). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.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 ICs of Recognition Category R or F.BNP Basis Reference(s):

1. BNP Technical Specifications Definitions section 1.1 2. BNP Technical Specifications Bases 3.4.5 3. BNP UFSAR section 5.1 Reactor Coolant System and Connected Systems 4. NEI 99-01 SU4 I OPEP-02.2.1 Rev. 6 1 Page 206 of 303 ATTACHMENT 1 Page 181 of 210 EAL Bases Category: Subcategory:

S -System Malfunction 6 -RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL: SU6.1 Unusual Event An automatic scram did not reduce reactor power to < 2% (APRM downscale) after any RPS setpoint is exceeded AND A subsequent automatic scram or manual scram action taken at the reactor control console (Manual PBs, Mode Switch, ARI) is successful in shutting down the reactor as indicated by reactor power < 2% (APRM downscale) (Note 8)Note 8: A manual scram 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. Mode Applicability: 1 -Power Operations, 2 -Startup Definition(s): None BNP Basis: The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protection System (RPS) scram function. A reactor scram is automatically initiated by the Reactor Protection System (RPS) when certain continuously monitored parameters exceed predetermined setpoints (ref. 1).Following a successful reactor scram, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative period. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-scram response from an automatic reactor scram signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful scram has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the APRM downscale setpoint of 2%.OPEP-02.2.1 Rev. 6 Page 207 of 303 ATTACHMENT 1 Page 182 of 210 EAL Bases For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., manual scram pushbuttons, mode switch, or ARI actuation). Reactor shutdown achieved by use of the OEOP-01-LEP-02 actions does not constitute a successful manual scram (ref. 4).Following any automatic RPS scram signal, OEOP-01 (ref. 2) and OEOP-01-LPC (ref. 3)prescribe insertion of redundant manual scram signals to back up the automatic RPS scram function and ensure reactor shutdown is achieved. Even if the first subsequent manual scram signal inserts all control rods to the full-in position immediately after the initial failure of the automatic scram, the lowest level of classification that must be declared is an Unusual Event.Taking the mode switch to shutdown is a manual scram action. When the Mode Switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered.If reactor power remains above the lowered setpoint, an automatic scram is initiated. For the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power below 2% is not considered a successful automatic scram. If automatic actuation of ARI has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. However, a successful automatic or manual initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions.In the event that the operator identifies a reactor scram is imminent and initiates a successful manual reactor scram before the automatic scram setpoint is reached, no declaration is required. The successful manual scram of the reactor before it reaches its automatic scram setpoint or reactor scram signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. If manual reactor scram actions fail to reduce reactor power below 2% (ref. 2, 3), the event escalates to the Alert under EAL SA6.1.If by procedure, operator actions include the initiation of an immediate manual scram following receipt of an automatic scram signal and there are no clear indications that the automatic scram failed (such as a time delay following indications that a scram setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic scram or manual actions. If a subsequent review of the scram actuation indications reveals that the automatic scram did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 50.72 should be considered for the transient event.I OPEP-02.2.1 Rev. 6 1 Page 208 of 303 ATTACHMENT 1 Page 183 of 210 EAL Bases NEI 99-01 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 reactor control consoles 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 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 reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram) 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 reactor control consoles 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". 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 reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FAI. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.I OPEP-02.2.1 Rev. 6 1 Page 209 of 303 ATTACHMENT 1 Page 184 of 210 EAL Bases 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. BNP Basis Reference(s):

1. BNP Technical Specifications section 3.3.1.1 RPS Instrumentation

2. 1(2) EOP-01, Reactor Scram Procedure 3. 1(2) EOP-01-LPC, Level/Power Control 4. OEOP-01-LEP-02 Alternate Control Rod Insertion 5. NEI 99-01 SU5 I OPEP-02.2.1 I Rev. 6 Page 210 of 303 ATTACHMENT 1 Page 185 of 210 EAL Bases Category: Subcategory:

S -System Malfunction 6 -RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL: SU6.2 Unusual Event A manual scram did not reduce reactor power to < 2% (APRM downscale) after any manual scram action was initiated AND A subsequent automatic scram or manual scram action taken at the reactor control console (Manual PBs, Mode Switch, ARI) is successful in shutting down the reactor as indicated by reactor power < 2% (APRM downscale) (Note 8)Note 8: A manual scram 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. Mode Applicability: 1 -Power Operations, 2 -Startup Definition(s): None BNP Basis: This EAL addresses a failure of a manually initiated scram in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual scram is successful in shutting down the reactor (reactor power < 2%). (ref. 1).Following a successful reactor scram, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative period. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-scram response from a manual reactor scram signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful scram has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the APRM downscale setpoint of 2%.OPEP-02.2.1 Rev. 6 Page 211 of 303 ATTACHMENT 1 Page 186 of 210 EAL Bases For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., manual scram pushbuttons, mode switch, or ARI actuation). Reactor shutdown achieved by use of the 0EOP-01-LEP-02 actions does not constitute a successful manual scram (ref. 2, 3).Taking the mode switch to shutdown is a manual scram action. When the Mode Switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered.If reactor power remains above the lowered setpoint, an automatic scram is initiated. Successful automatic or manual initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions.If both subsequent automatic and subsequent manual reactor scram actions in the Control Room fail to reduce reactor power below the power associated with the safety system design (< 2%) following a failure of an initial manual scram, the event escalates to an Alert under EAL SA6.1 NEI 99-01 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 reactor control consoles 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 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 reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram) 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.I OPEP-02.2.1 Rev. 6 Page 212 of 303 ATTACHMENT 1 Page 187 of 210 EAL Bases A manual action at the reactor control consoles 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". 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 reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.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.

BNP Basis Reference(s):

1. BNP Technical Specifications section 3.3.1.1 RPS Instrumentation

2. 1(2) EOP-01, Reactor Scram Procedure 3. OEOP-01-LEP-02 Alternate Control Rod Insertion 4. NEI 99-01 SU5 I OPEP-02.2.1 I Rev. 6 Page 213 of 303 ATTACHMENT 1 Page 188 of 210 EAL Bases Category: Subcategory:

Initiating Condition: S -System Malfunction 2 -RPS Failure Automatic or manual scram fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL: SA6.1 Alert An automatic or manual scram fails to reduce reactor power to < 2% (APRM downscale) AND Manual scram actions taken at the reactor control console (Manual PBs, Mode Switch, ARI) are not successful in shutting down the reactor as indicated by reactor power > 2%(Note 8)Note 8: A manual scram 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. Mode Applicability: 1 -Power Operations, 2 -Startup Definition(s): None BNP Basis: This EAL addresses any automatic or manual reactor scram signal that fails to shut down the reactor followed by a subsequent manual scram that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed.For the purposes of emergency classification at the Alert level, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., manual scram pushbuttons, mode switch, or ARI actuation). Reactor shutdown achieved by use of the OEOP-01-LEP-02 actions does not constitute a successful manual scram (ref. 1).IOPEP-02.2.1 I Rev. 6 1 Page 214 of 303 ATTACHMENT 1 Page 189 of 210 EAL Bases For the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power below 2% is not considered a successful automatic scram. If automatic actuation of ARI has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. However, a successful automatic initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions.The APRM downscale trip setpoint (2%) is a minimum reading on the power range scale that indicates power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM)indications or other reactor parameters (steam flow, RPV pressure, torus temperature trend)can be used to determine if reactor power is greater than 2 % power (ref. 2, 3).Escalation of this event to a Site Area Emergency would be under EAL SS6.NEI 99-01 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 subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.A manual action at the reactor control console 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. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers). Actions taken at backpanels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control console".Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.I OPEP-02.2.1 Rev. 6 Page 215 of 303 ATTACHMENT 1 Page 190 of 210 EAL Bases 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 shut down the reactor is prolonged enough to cause a challenge to RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS6 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. BNP Basis Reference(s):

1. OEOP-01-LEP-02, Alternate Control Rod Insertion 2. 1(2) EOP-01, Reactor Scram Procedure 3. 1(2) EOP-01-LPC, Level/Power Control 4. NEI 99-01 SA5 I OPEP-02.2.1 Rev. 6 Page 216 of 303 ATTACHMENT 1 Page 191 of 210 EAL Bases Category: Subcategory:

S -System Malfunction 2 -RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to RPV water level or RCS heat removal EAL: SS6.1 Site Area Emergency An automatic or manual scram fails to reduce reactor power to < 2% (APRM downscale) AND All actions to shut down the reactor are not successful as indicated by reactor power> 2%AND EITHER: " RPV level cannot be restored and maintained > LL-4 or cannot be determined" Suppression pool water temperature and RPV pressure cannot be maintained below the HCTL Mode Applicability: 1 -Power Operations, 2 -Startup Definition(s): None BNP Basis: This EAL addresses the following: " Any automatic reactor scram signal followed by a manual scram that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed (EAL SA6.1), and* Indications that either core cooling is extremely challenged or heat removal is extremely challenged. I OPEP-02.2.1 Rev. 6 1 Page 217 of 303 ATTACHMENT 1 Page 192 of 210 EAL Bases Reactor shutdown achieved by use of OEOP-01-LEP-02 Alternate Control Rod Insertion is also credited as a successful manual scram provided reactor power can be reduced below the APRM downscale trip setpoint before indications of an extreme challenge to either core cooling or heat removal exist.The APRM downscale trip setpoint (2%) is a minimum reading on the power range scale that indicates power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM)indications or other reactor parameters (steam flow, RPV pressure, torus temperature trend)can be used to determine if reactor power is greater than 2% power (ref. 1, 2).The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS barriers.Indication that core cooling is extremely challenged is manifested by inability to restore and maintain RPV water level above LL-4. LL-4 is the Minimum Steam Cooling RPV Water Level (MSCRWL) (ref. 3). The MSCRWL is the lowest RPV level at which the covered portion of the reactor core will generate sufficient steam to prevent any clad temperature in the uncovered part of the core from exceeding 1500'F. This water level is utilized in the EOPs to preclude fuel damage when RPV level is below the top of active fuel. RPV level below the MSCRWL for an extended period of time without satisfactory core spray cooling could be a precursor of a core melt sequence. When RPV level cannot be determined, EOPs require entry to EOP-01-RXFP, Reactor Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained. When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EOP-01-RXFP specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Alternate Flooding Pressure (ref. 4).The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool water temperature from which Emergency RPV Depressurization will not raise suppression chamber pressure above Primary Containment Pressure Limit A (PCPL-A), while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.The HCTL is a function of RPV pressure and torus water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant. This threshold is met when the final step SP/T-1 3 of section SP/T in EOP-02-PCCP, Primary Containment Control, is reached (ref. 5). This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature. I OPEP-02.2.1 Rev. 6 1 Page 218 of 303 ATTACHMENT 1 Page 193 of 210 EAL Bases NEI 99-01 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, 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 shut down 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.Escalation of the emergency classification level would be via IC RG1 or FG1.BNP Basis Reference(s):

1. EOP-01 Reactor Scram Procedure 2. EOP-01 -LPC Level/Power Control 3. OEOP-01-NL EOP-SAMG Numerical Limits and Values, Attachment 1, pg 37-40, Figures 1-10 and 1-11 4. EOP-01-RXFP, Reactor Flooding 5. EOP-02-PCCP, Primary Containment Control 6. NEI 99-01 SS5 I OPEP-02.2.1 Rev. 6 1 Page 219 of 303 ATTACHMENT 1 Page 194 of 210 EAL Bases Category: Subcategory:

Initiating Condition: EAL: S -System Malfunction 7 -Loss of Communications Loss of all onsite or offsite communications capabilities SU7.1 Unusual Event Loss of all Table S-3 onsite communication methods OR Loss of all Table S-3 offsite communication methods OR Loss of all Table S-3 NRC communication methods Table S-3 Communication Methods System Onsite Offsite NRC Public Address System X PBX Telephone System X X X Corporate Telephone X X X Communications System Commercial Telephones X X X Satellite Phones X X Cellular Phones X X NRC Emergency X Telecommunications System IOPEP-02.2.1 I Rev. 6 Page 220 of 303 ATTACHMENT 1 Page 195 of 210 EAL Bases Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: Onsite/offsite/NRC communications include one or more of the systems listed in Table S-3 (ref.1).Public Address System The Brunswick Plant public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature. This system is powered from the plant uninterruptible power supply which employs battery reserve as well as diesel generator emergency supply.PBX Telephone System The Brunswick Site PBX telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code. The PBX telephone system also provides for outside communications. The PBX switch located in the TSC/EOF building is also backed up by a battery UPS capable of supplying power for a minimum of 8 hours and is augmented by a Diesel Generator capable of supplying power to the TSC/EOF building for at least 5 days.Corporate Telephone Communications System (Voicenet and/or DEMNET)Interconnected through the site PBX and the emergency telephone system, the Voicenet system provides a means to communicate with other corporate locations with which the plant has a need to communicate. This system bypasses external commercial telephone lines and switching equipment. Corporate transmission facilities provide fiber optic, copper-wire, and microwave radio to ensure a high degree of system reliability. In addition to the redundancy provided by the three system options, backup power is provided for the systems. DEMNET is the primary means of offsite communication. This circuit allows intercommunication among the EOF, TSC, control room, counties, and states. DEMNET operates as an internet based (VoIP)communications system with a satellite back-up. Should the internet transfer rate become slow or unavailable, the DEMNET will automatically transfer to satellite mode.I OPEP-02.2.1 Rev. 6 1 Page 221 of 303 ATTACHMENT 1 Page 196 of 210 EAL Bases Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by Duke Energy in four ways: (1) tie-ins through the PBX to any other plant location, (2) lines to plant emergency facilities, (3) lines to the Joint Information Center for public information purposes, and (4) lines to the AEF. The local service provider provides primary and secondary power for their lines at the Central Office.Satellite Phones A total of three portable satellite telephones are available which enable communication when all other phone systems are inoperable, e.g. following a major hurricane. These portable systems can be powered by internal batteries, external DC sources as well as external AC sources. Two of these phones require outside use, while one phone may used either outside or in the EOF with a permanently mounted external antenna.Cellular Phones Selected plant personnel are provided with cellular telephones. These phones may be used during emergencies if other communications means are not readily available or are inoperable. These phones are not expected to be used in the Control Room or Power Block due to interference with plant equipment and loss of signal to the phone.NRC Emergency Telecommunications System The NRC uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRC regional and national offices. The Duke Energy communications line provides a link independent of the local public telephone network.Telephones connected to this network are located in the Brunswick Control Room, Technical Support Center, and Emergency Operations Facility. These lines will not function if the PBX Telephone System fails.This EAL is the hot condition equivalent of the cold condition EAL CU5.1.NEI 99-01 Basis: This IC addresses a significant loss of onsite 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 onsite information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).OPEP-02.2.1 Rev. 6 Page 222 of 303 ATTACHMENT 1 Page 197 of 210 EAL Bases The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations. The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the State, Brunswick and New Hanover County EOCs The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. BNP Basis Reference(s):

1. OERP Radiological Emergency Response Plan Appendix A 2. SD-48 Communication Systems 3. NEI 99-01 SU6 I OPEP-02.2.1 Rev. 6 Page 223 of 303 ATTACHMENT 1 Page 198 of 210 EAL Bases Category:

S -System Malfunction Subcategory: 8 -Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL: SA8.1 Alert The occurrence of any Table S-4 hazardous event AND EITHER: " 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 Table S-4 Hazardous Events* Seismic event (earthquake)

Internal or external FLOODING event* High winds or tornado strike* FIRE* EXPLOSION* Other events with similar hazard characteristics as determined by the Shift Manager Mode Applicability:

1 -Power Operations, 2 -Startup, 3 -Hot Shutdown I OPEP-02.2.1 Rev. 6 Page 224 of 303 ATTACHMENT 1 Page 199 of 210 EAL Bases Definition(s): 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 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.FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.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 (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. 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. BNP Basis: " The significance of seismic events are discussed under EAL HU2.1 (ref. 1)." Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 2, 3).* Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 135 mph. (ref. 4).I OPEP-02.2.1 Rev. 6 1 Page 225 of 303 ATTACHMENT 1 Page 200 of 210 EAL Bases" Areas containing functions and systems required for safe shutdown of the plant are identified by Fire Zone in the fire response procedure (ref. 5, 6)." An explosion that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL.NEI 99-01 Basis: This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode.This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.The first condition 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 condition 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 FS1 or RSI.BNP Basis Reference(s):

1. 1(2)APP-UA-28 6-4 Seismic Event 2. OAOP-13.0 Operations During Hurricane, Flood Conditions, Tornado or Earthquake

3. Updated FSAR section 3.4.2 Protection From Internal Flooding 4. Updated FSAR Section 2.3.1.2.7 5. BNP-E-9.004 Safe Shutdown Analysis Report 6. OPFP-PBAA Power Block Auxiliary Areas Prefire Plan 7. NEI 99-01 SA9 I OPEP-02.2.1 I Rev. 6 1 Page 226 of 303 ATTACHMENT 1 Page 201 of 210 EAL Bases Category E -Independent Spent Fuel Storage Installation (ISFSI)EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold)An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a cask/canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel. Formal offsite planning is not required because the postulated worst-case accident involving an ISFSI has insignificant consequences to the public health and safety.An Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated.A hostile security event that leads to a potential loss in the level of safety of the ISFSI is a classifiable event under Security category EAL HS1.1.Minor surface damage that does not affect storage cask/canister boundary is excluded from the scope of these EALs.I OPEP-02.2.1 Rev. 6 1 Page 227 of 303 ATTACHMENT 1 Page 202 of 210 EAL Bases Category:

E -ISFSI Sub-category: None Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL: EUI.1 Notification of Unusual Event Damage to a loaded canister confinement boundary as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask > any of the following:

1,400 mrem/hr on the HSM-H front surface* 10 mrem/hr on the HSM-H door centerline
20 mrem/hr on the end shield wall exterior Mode Applicability:

All Definition(s): CONFINEMENT BOUNDARY-. The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the BNP ISFSI, Confinement Boundary is defined as the Dry Shielded Canister (DSC).BNP Basis: The BNP ISFSI utilizes the NUHOMS Type 2 -61 BTH dry spent fuel storage (ref. 1, 2).The NUHOMS Type 2 61 BTH spent fuel storage system is a modular canister based spent fuel storage and transfer system and consists of the following components: " A 61 BTH Dry Shielded Canister (DSC) provides confinement, an inert environment, structural support, and criticality control for 61 BWR fuel assemblies." A horizontal storage module (HSM-H) is provided for environmental protection, shielding, and heat rejection during storage." An OS1 97FC-B transfer cask that supports onsite transfer of the 61 BTH DSC.I OPEP-02.2.1 Rev. 6 Page 228 of 303 ATTACHMENT 1 Page 203 of 210 EAL Bases The NUHOMS System confinement vessel is the DSC. The DSC is welded and designed to provide confinement of all radionuclides under normal, off-normal, and accident conditions. Confinement boundary is defined as the barrier(s) between areas containing radioactive substances and the environment. Therefore, damage to a confinement boundary must be a confirmed physical breach between the spent fuel and the environment for the Dry Shield Canister (DSC).The values shown represent 2 times the limits specified in the ISFSI Certificate of Compliance (COC) Technical Specification for radiation external to a loaded (NUHOMS Type 2 -61 BTH)MPC (HSM-H) overpack (ref. 1, 2). The survey method(s) used to assess this EAL threshold shall be consistent with those used to ensure compliance with the COC Technical Specification limits (ref. 2).NEI 99-01 Basis: This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.The existence of "damage" is determined by radiological survey. The technical specification multiple of"2 times", which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.Security-related events for ISFSs are covered under ICs HU1 and HAl.I OPEP-02.2.1 Rev. 6 1 Page 229 of 303 ATTACHMENT 1 Page 204 of 210 EAL Bases BNP Basis Reference(s):

1. OPLP-36 BNP 10CFR50.72.212 Report 2. Transnuclear, Inc. Standardized NUHOMS Horizontal Modular Storage System Certificate of Compliance No. 1004, Ammendment 10 Enclosure 1 3. NGGM-PM-0028 Transnuclear NUHOMS Dry Fuel Storage Program Manual 4. NEI 99-01 E-HU1 I 0PEP-02.2.1 Rev. 6 Page 230 of 303 ATTACHMENT 1 Page 205 of 210 EAL Bases Category F -Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature

> 212 0 F); EALs in this category are applicable only in one or more hot operating modes.EALs in this category represent threats to the defense in depth design concept that precludes the release of highly 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: A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.B. Reactor Coolant System (RCS): 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.C. Containment (PC): The drywell, the torus, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves comprise the PC barrier. Primary Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency. The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1 (Attachment 2). "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials. "Potential Loss" means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level: Alert: Any loss or any potential loss of either Fuel Clad or RCS Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of third barrier I OPEP-02.2.1 Rev. 6 Page 231 of 303 1 ATTACHMENT 1 Page 206 of 210 EAL Bases The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations: " The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Primary Containment Barrier.* Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.* For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC RG1 has been exceeded.* The fission product barrier thresholds specified within a scheme reflect plant-specific BNP design and operating characteristics." As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location-inside the primary containment, an interfacing system, or outside of the primary containment. The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage." At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the SEC would have more assurance that there was no immediate need to escalate to a General Emergency. OPEP-02.2.1 I Rev. 6 Page 232 of 303 ATTACHMENT 1 Page 207 of 210 EAL Bases Category: Subcategory: Fission Product Barrier Degradation N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS barrier EAL: FA1.1 Alert Any loss or any potential loss of either Fuel Clad or RCS (Table F-i)Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: Fuel Clad, RCS and Primary Containment comprise the fission product barriers. Table F-1 (Attachment

2) lists the fission product barrier thresholds, bases and references.

At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.1 NEI 99-01 Basis: None BNP Basis Reference(s):

1. NEI 99-01 FAI IOPEP-02.2.1 I Rev. 6 Page 233 of 303 ATTACHMENT 1 Page 208 of 210 EAL Bases Category:

Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL: FSI.1 Site Area Emergency Loss or potential loss of any two barriers (Table F-i)Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment

2) lists the fission product barrier thresholds, bases and references.

At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions: , One barrier loss and a second barrier loss (i.e., loss -loss)* One barrier loss and a second barrier potential loss (i.e., loss -potential loss)* One barrier potential loss and a second barrier potential loss (i.e., potential loss -potential loss)IOPEP-02.2.1 I Rev. 6 Page 234 of 303 ATTACHMENT 1 Page 209 of 210 EAL Bases At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Site Emergency Coordinator would have greater assurance that escalation to a General Emergency is less imminent.NEI 99-01 Basis: None BNP Basis Reference(s):

1. NEI 99-01 FS1 I OPEP-02.2.1 I Rev. 6 Page 235 of 303 ATTACHMENT 1 Page 210 of 210 EAL Bases Category: Subcategory:

Fission Product Barrier Degradation N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL: FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of third barrier (Table F-I)Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Shutdown Definition(s): None BNP Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment

2) lists the fission product barrier thresholds, bases and references.

At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:

Loss of Fuel Clad, RCS and Containment barriers* Loss of Fuel Clad and RCS barriers with potential loss of Containment barrier* Loss of RCS and Containment barriers with potential loss of Fuel Clad barrier* Loss of Fuel Clad and Containment barriers with potential loss of RCS barrier NEI 99-01 Basis: None BNP Basis Reference(s):

1. NEI 99-01 FG1 OPEP-02.2.1 Rev. 6 Page 236 of 303 ATTACHMENT 2 Page 1 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Introduction Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment).

The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds. The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are: A. RPV Water Level B. RCS Leak Rate B. Primary Containment Conditions C. Primary Containment Radiation/RCS Activity D. Primary Containment Integrity or Bypass E. SEC Jugement Each category occupies a row in Table F-1 thus forming a matrix defined by the categories. The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell.Thresholds are assigned sequential numbers within each Loss and Potential Loss column beginning with number one. In this manner, a threshold can be identified by its category title and number. For example, the first Fuel Clad barrier Loss in Category A would be assigned"FC Loss A.1 ," the third Containment barrier Potential Loss would be assigned "PC P-Loss B.3," etc.If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers.IOPEP-02.2.1 I Rev. 6 Page 237 of 303 ATTACHMENT 2 Page 2 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-i, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost -even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS barriers and a Potential Loss of the Containment barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FS1.1, and FA1.1 to determine the appropriate emergency classification. In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the RCS barrier and finally the Containment barrier threshold bases. In each barrier, the bases are given according category Loss followed by category Potential Loss beginning with Category A, then B,..., F.OPEP-02.2.1 Rev. 6 Page 238 of 303 ATTACHMENT 2 Page 3 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier Reactor Coolant System Barrier Primary Containment Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A 1. RPV level cannot be restored and 1. RPV level cannot be restored and 1. Entry to SAMG-01 required maintained > TAF or cannot be maintained > TAF or cannot be None None 1. Entry to SAMG-01 required RPV Water determined determined Level 1. UNISOLABLE primary system 1. UNISOLABLE pdmary system leakage that results in exceeding leakage that results in exceeding 1. UNISOLABLE break in any of the EITHER of the following: EITHER of the following: following: One or more Secondary

One or more Secondary* Main steam Containment area radiation Containment area radiation* HPCI steam Line Maximum Normal Operating Maximum Safe Operating B None None
RCIC steam Line Limits (OEOP-03-SCCP Table Limits (OEOP-03-SCCP Table None RCS Leak Rate
RWCU 3) 3)* Feedwater One or more Secondary One or more Secondary 2. Emergency Depressurization is Containment area temperature Containment area required Maximum Normal Operating temperature Maximum Safe Limits Operating Limits (OEOP-03-SCCP Table 1) (OEOP-03-SCCP Table 1)1. UNPLANNED rapid drop in 1. Pnmary Containment pressure Primary Containment pressure > 62 psig C following Primary Containment None None 1. Primary Containment pressure>

None pressure rise 2. Deflagration concentrations exist PC 1.7 psig due to RCS leakage inside PC (H 2 6% AND 02 5%)Conditions

2. Primary Containment pressure 3. Heat Capacity Temperature Limit response not consistent with (HCTL) exceeded LOCA conditions D 1. Drywell radiation

> 2,000 Rthr PC Rad .None 1. Drywell radiation v 27 R/hr with None None 1. Drywell radiation > 20,000 R/hr RCS 2. Pnmary coolant activity > 300 reactor shutdown Activity pCilgm 1-131 dose equivalent

1. UNISOLABLE direct downstream pathway to the environment E exists after Primary Containment None None None None isolation signal None PC Integrity or Bypass 2. Intentional Primary Containment venting per EOPs F 1. Any condition in the opinion of 1. Any condition in the opinion of 1. Any condition in the opinion of 1. Any condition in the opinion of the 1. Any condition in the opinion of 1. Any condition in the opinion of the the SEC that indicates loss of the SEC that indicates potential the SEC that indicates loss of the SEC that indicates potential loss of the SEC that indicates loss of the SEC that indicates potential loss of ED the fuel clad barrier loss of the fuel clad barrier RCS barrer the RCS barrier Primary Containment barrier the Primary Containment barrier Judgment I OPEP-02.2.1 Rev. 6 Page 239 of 303 ATTACHMENT 2 Page 4 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

A. RPV Level Degradation Threat: Loss Threshold:

1. Entry to SAMG-01 required Definition(s):

N/A BNP Basis: EOP-01-RVCP, EOP-01-LPC and EOP-01-RXFP specify the requirement for entry to SAMG-01 when core cooling is severely challenged. These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined. SAMG-01 is required when (ref. 1):* Reactor water level cannot be restored and maintained above -57.5 inches (Jet Pump Suction) with at least one core spray pump injecting into the reactor vessel" Reactor vessel water level cannot be restored and maintained above LL-4 (MSCRWL)* The reactor vessel flooding conditions cannot be restored and maintained (5 SRVs open and reactor vessel pressure more than 50 psig above suppression chamber pressure)" When at least 1 SRV cannot be opened and reactor vessel pressure cannot be restored and maintained above the minimum alternate reactor vessel flooding pressure (Table 1 values that are dependent on number of open SRVs)The above EOP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad.This threshold is also a Potential Loss of the Containment barrier (PC P-Loss A.1). Since entry to SAMG-01 occurs after core uncovery has occurred a Loss of the RCS barrier exists (RCS Loss A.1). Entry to SAMG-01, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification. OPEP-02.2.1 Rev. 6 Page 240 of 303 ATTACHMENT 2 Page 5 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases NEI 99-01 Basis: The Loss threshold represents the EOP requirement for entry to SAMG-01. This is identified in the BWROG EPGs/SAGs when the phrase, "enter all Severe Accident Guidelines" appears.Since a site-specific RPV water level is not specified here, the Loss threshold phrase, "Entry to SAMG-01 required," also accommodates the EOP need to enter SAMG-01 when RPV water level cannot be determined and core damage due to inadequate core cooling is believed to be occurring. BNP Basis Reference(s):

1. OSAMG-06.0 SAMG Primary Containment Flooding Basis Document 2. NEI 99-01 RPV Water Level Fuel Clad Loss 2.A OPEP-02.2.1 Rev. 6 Page 241 of 303 ATTACHMENT 2 Page 6 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

A. RPV Level Degradation Threat: Potential Loss Threshold:

1. RPV level cannot be restored and maintained

> TAF or cannot be determined Definition(s): N/A BNP Basis: An RPV level instrument reading of -7.5 in. indicates RPV level is at the top of active fuel (TAF)(ref. 1). When RPV level is at or above the TAF, the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV level is below TAF, the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray cooling). If core uncovery is threatened, the EOPs specify alternate, more extreme, RPV level control measures in order to restore and maintain adequate core cooling. Since core uncovery begins if RPV level drops to TAF, the level is indicative of a challenge to core cooling and the Fuel Clad barrier.When RPV level cannot be determined, EOPs require entry to EOP-01-RXFP, Reactor Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained (ref. 2, 3). When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EOP-01-RXFP specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Alternate Reactor Vessel Flooding Pressure (in scram-failure events) (ref. 4). If RPV water level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists.Note that EOP-01-LPC, Level/Power Control, may require intentionally lowering RPV water level to TAF and control level between the LL-4, the Minimum Steam Cooling RPV Water Level (MSCRWL) and TAF (ref. 3). Under these conditions, a high-power ATWS event exists and requires at least an Alert classification in accordance with the System Malfunction -RPS Failure EALs, however under these conditions a potential loss of the fuel clad does not exist.IOPEP-02.2.1 I Rev. 6 Page 242 of 303 ATTACHMENT 2 Page 7 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases NEI 99-01 Basis: 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 1 .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 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. 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 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 SA6 or SS6 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 loss of the fuel clad barrier is specified. IOPEP-02.2.1 I Rev. 6 1 Page 243 of 303 ATTACHMENT 2 Page 8 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases BNP Basis Reference(s):

1. OEOP-01-NL EOP-SAMG Numerical Limits and Values, Attachment 1 2. EOP-01-RVCP Reactor Vessel Control 3. EOP-01-LPC, Level/Power Control 4. EOP-01-RXFP, Reactor Flooding 5. NEI 99-01 RPV Water Level Fuel Clad Potential Loss 2.A I 0PEP-02.2.1 Rev. 6 Page 244 of 303 ATTACHMENT 2 Page 9 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

B. RCS Leak Rate Degradation Threat: Loss Threshold: None I OPEP-02.2.1 Rev. 6 Page 245 of 303 ATTACHMENT 2 Page 10 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. RCS Leak Rate Degradation Threat: Potential Loss Threshold: None IOPEP-02.2.1 I Rev. 6 Page 246 of 303 ATTACHMENT 2 Page 11 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. PC Conditions Degradation Threat: Loss Threshold: None I OPEP-02.2.1 Rev. 6 Page 247 of 303 ATTACHMENT 2 Page 12 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. PC Conditions Degradation Threat: Potential Loss Threshold: None I OPEP-02.2.1 Rev. 6 1 Page 248 of 303 ATTACHMENT 2 Page 13 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. PC Radiation / RCS Activity Degradation Threat: Loss Threshold:

1. Drywell radiation

> 2,000 R/hr Definition(s): None BNP Basis: The Drywell High-Range Radiation Monitor (1 (2)D22-RI-4195, 1 (2)D22-RI-4196, 1 (2)D22-RI-4197, 1(2)D22-RI-4198) reading of 2,000 R/hr indicates the release of reactor coolant into the drywell with elevated activity indicative of fuel damage.Based on 2% clad damage, a containment radiation level of 2000 R/hr is derived as follows: Per OPEP-03.6.3 Table 3, 100% Cladding Damage column 'No Spray' for 1 hour after shutdown is 100,000 R/hr. Per Step 7.2.2.1, 0.02 x 100,000 R/hr = 2000 R/hr containment radiation (ref. 1).NEI 99-01 Basis: 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 dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage.Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold D.1 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. I OPEP-02.2.1 I Rev. 6 1 Page 249 of 303 ATTACHMENT 2 Page 14 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases BNP Basis Reference(s):

1. OPEP-03.6.3 Estimate of the Extent of Core Damage Under Accident Conditions

2. NEI 99-01 Primary Containment Radiation Fuel Clad Loss 4.A I OPEP-02.2.1 Rev. 6 1 Page 250 of303 ATTACHMENT 2 Page 15 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Fuel Clad Barrier: Category: D. PC Radiation

/ RCS Activity Degradation Threat: Loss Threshold:

2. Primary coolant activity > 300 pCi/gm 1-131 dose equivalent BNP Basis: None NEI 99-01 Basis: This threshold indicates that RCS radioactivity concentration is greater than 300 gCi/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.There is no Potential Loss threshold associated with Primary Containment Radiation.

BNP Basis Reference(s):

1. NEI 99-01 RCS Activity Fuel Clad Loss 1.A I OPEP-02.2.1 Rev. 6 Page 251 of 303 ATTACHMENT 2 Page 16 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

D. PC Radiation / RCS Activity Degradation Threat: Potential Loss Threshold: None I OPEP-02.2.1 Rev. 6 Page 252 of 303 ATTACHMENT 2 Page 17 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. PC Integrity or Bypass Degradation Threat: Loss Threshold: None OPEP-02.2.1 Rev. 6 Page 253 of 303 ATTACHMENT 2 Page 18 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. PC Integrity or Bypass Degradation Threat: Potential Loss Threshold: None I OPEP-02.2.1 Rev. 6 Page 254 of 303 I ATTACHMENT 2 Page 19 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: F. SEC Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Site Emergency Coordinator that indicates loss of the Fuel Clad barrier Definition(s):

None BNP Basis: The Site Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks." Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Site Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that are to be used by the Site Emergency Coordinator in determining whether the Fuel Clad barrier is lost BNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A OPEP-02.2.1 Rev. 6 Page 255 of 303 ATTACHMENT 2 Page 20 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

F. SEC Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Site Emergency Coordinator that indicates potential loss of the Fuel Clad barrier BNP Basis: The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Site Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that are to be used by the Site Emergency Coordinator in determining whether the Fuel Clad barrier is potentially lost. The Site Emergency Coordinator should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. BNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A IOPEP-02.2.1 Rev. 6 Page 256 of 303 ATTACHMENT 2 Page 21 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

A. RPV Water Level Degradation Threat: Loss Threshold:

1. RPV level cannot be restored and maintained

> TAF or cannot be determined Definition(s): None BNP Basis: An RPV level instrument reading of -7.5 in. indicates RPV level is at the top of active fuel (TAF)(ref. 1). TAF is significantly lower than the normal operating RPV level control band. To reach this level, RPV inventory loss would have previously required isolation of the RCS and Primary Containment barriers, and initiation of all ECCS. If RPV level cannot be maintained above TAF, ECCS and other sources of RPV injection have been ineffective or incapable of reversing the decreasing level trend. The cause of the loss of RPV inventory is therefore assumed to be a LOCA. By definition, a LOCA event is a Loss of the RCS barrier.When RPV level cannot be determined, EOPs require entry to EOP-01-RXFP, Reactor Flooding (ref. 2). The instructions in EOP-01-RXFP specify emergency depressurization of the RPV, which is defined to be a Loss of the RCS barrier (RCS Loss C.4).The conditions of this threshold are also a Potential Loss of the Fuel Clad barrier (FC P-Loss A.1). A Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier requires a Site Area Emergency classification. Note that EOP-01-LPC, Level/Power Control, may require intentionally lowering RPV water level to TAF and control level between LL-4, the Minimum Steam Cooling RPV Water Level (MSCRWL), and TAF (ref. 3). Under these conditions, a high-power ATWS event exists and requires at least an Alert classification in accordance with the System Malfunction -RPS Failure EALs, however under these conditions a loss of the RCS does not exist.IOPEP-02.2.1 I Rev. 6 Page 257 of 303 ATTACHMENT 2 Page 22 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases NEI 99-01 Basis: 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 1 .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 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. BNP Basis Reference(s):

1. OEOP-01-NL EOP-SAMG Numerical Limits and Values, Attachment 1 2. EOP-01-RVCP Reactor Vessel Control 3. EOP-01-LPC, Level/Power Control 4. EOP-01-RXFP, Reactor Flooding 5. NEI 99-01 RPV Water Level RCS Loss 2.A IOPEP-02.2.1 I Rev. 6 Page 258 of 303 ATTACHMENT 2 Page 23 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

A. RPV Water Level Degradation Threat: Potential Loss Threshold: None I OPEP-02.2.1 Rev. 6 Page 259 of 303 ATTACHMENT 2 Page 24 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:

1. UNISOLABLE break outside Primary Containment in any of the following:

Main steam line" HPCI steam line" RCIC steam line* RWCU" Feedwater Definition(s):

UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.BNP Basis: The conditions of this threshold include required containment isolation failures allowing a flow path to the environment. A release pathway outside primary containment exists when flow is not prevented by downstream isolations. In the case of a failure of both isolation valves to close but in which no downstream flowpath exists, emergency declaration under this threshold would not be required. Similarly, if the emergency response requires the normal process flow of a system outside primary containment (e.g., EOP requirement to bypass MSIV low RPV water level interlocks and maintain the main condenser as a heat sink using main turbine bypass valves), the threshold is not met. The combination of these threshold conditions represent the loss of both the RCS and Primary Containment (see PC Loss E.1) barriers and justifies declaration of a Site Area Emergency (i.e., Loss or Potential Loss of any two barriers). Even though RWCU and Feedwater systems do not contain steam, they are included in the list because an unisolable break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume systems directly connected to the RCS.NEI 99-01 Basis: 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.°PEP-02.2.1 Rev. 6 Page 260 of 303 ATTACHMENT 2 Page 25 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases BNP Basis Reference(s):

1. 1(2)OP-01 Nuclear Boiler System 2. 1(2)OP-25 Main Steam System Operating Procedure 3. 1(2)OP-19 High Pressure Coolant Injection System Operating Procedure 4. 1(2)OP-16 Reactor Core Isolation Cooling System Operating Procedure 5. 1(2)OP-1 4 Reactor Water Cleanup System Operating Procedure 6. 1(2)OP-32 Condensate and Feedwater System Operating Procedure 7. NEI 99-01 RCS Leak Rate RCS Loss 3.A I OPEP-02.2.1 I Rev. 6 Page 261 of 303 ATTACHMENT 2 Page 26 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

B. RCS Leak Rate Degradation Threat: Loss Threshold:

2. Emergency Depressurization is required Definition(s):

N/A BNP Basis: Plant symptoms requiring Emergency Depressurization per the EOPs are indicative of a loss of the RCS barrier. If Emergency depressurization is required, the plant operators are directed to open safety relief valves (SRVs) and keep them open regardless of any subsequent radiological release rate (ref. 1 -6). Even though the RCS is being vented into the suppression pool, a loss of the RCS should be considered to exist due to the diminished effectiveness of the RCS pressure barrier to a release of fission products beyond its boundary.NEI 99-01 Basis: 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) 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.BNP Basis Reference(s):

1. EOP-01-UG User's Guide 2. EOP-01-RVCP Reactor Vessel Control 3. EOP-01-LCP Level/Power Control 4. EOP-02-PCCP Primary Containment Control 5. OEOP-03-SCCP Secondary Containment Control 6. EOP-04-RRCP Radioactivity Release Control 7. EOP-01-RXFP Reactor Flooding 8. NEI 99-01 RCS Leak Rate RCS Loss 3.B OPEP-02.2.1 Rev. 6 Page 262 of 303 ATTACHMENT 2 Page 27 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Reactor Coolant System Barrier: Category: B. RCS Leak Rate Degradation Threat: Potential Loss Threshold:

1. UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

One or more Secondary Containment area radiation Maximum Normal Operating Limits (OEOP-03-SCCP Table 3)" One or more Secondary Containment area temperature Maximum Normal Operating Limits (OEOP-03-SCCP Table 1)Definition(s):

UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.BNP Basis: The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of unisolable primary system leakage outside the primary containment. The Maximum Normal Operating Limit values define this RCS threshold because they are the maximum normal operating values and signify the onset of abnormal system operation. When parameters reach this level, equipment failure or misoperation may be occurring. Elevated parameters may also adversely affect the ability to gain access to or operate equipment within the affected area. The locations into which the primary system discharge is of concern correspond to the areas addressed in OEOP-03-SCCP, Secondary Containment Control Tables (ref. 1).In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g.room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment. OPEP-02.2.1 Rev. 6 Page 263 of 303 ATTACHMENT 2 Page 28 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases NEI 99-01 Basis: 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.BNP Basis Reference(s):

1. OEOP-03-SCCP, Secondary Containment Control 2. NEI 99-01 RCS Leak Rate RCS Potential Loss 3.A OPEP-02.2.1 Rev. 6 Page 264 of 303 ATTACHMENT 2 Page 29 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

C. PC Conditions Degradation Threat: Loss Threshold:

1. Primary Containment pressure > 1.7 psig due to RCS leakage Definition(s):

None BNP Basis: The drywell high pressure scram setpoint is an entry condition to EOP-01-RVCP Reactor Vessel Control, and EOP-02-PCCP, Primary Containment Control (ref. 1, 2, 3). Normal primary containment pressure control functions (e.g., operation of drywell coolers, vent through SBGT, etc.) are specified in EOP-02-PCCP in advance of less desirable but more effective functions (e.g., operation of drywell or suppression pool sprays, etc.).In the BNP design basis, primary containment pressures above the drywell high pressure scram setpoint are assumed to be the result of a high-energy release into the containment for which normal pressure control systems are inadequate or incapable of reversing the increasing pressure trend. Pressures of this magnitude, however, can be caused by non-LOCA events such as a loss of drywell cooling or inability to control primary containment vent/purge (ref. 4).The threshold phrase "...due to RCS leakage" focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. PC pressure greater than 1.7 psig with corollary indications (e.g., drywell temperature, indications of loss of RCS inventory) should therefore be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressure greater than 1.7 psig should not be considered an RCS barrier Loss.NEI 99-01 Basis: 1.7 psig 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.IOPEP-02.2.1 I Rev. 6 Page 265 of 303 ATTACHMENT 2 Page 30 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases BNP Basis Reference(s):

1. OEOP-01-NL EOP-SAMG Numerical Limits and Values, Attachment 3 2. EOP-01-RVCP Reactor Vessel Control 3. EOP-02-PCCP Primary Containment Control 4. BNP Updated FSAR Chapter 6 Emergency Core Cooling Systems 5. NEI 99-01 Primary Containment Pressure RCS Loss I.A I OPEP-02.2.1 Rev. 6 Page 266 of 303 ATTACHMENT 2 Page 31 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

C. PC Conditions Degradation Threat: Potential Loss Threshold: None IOPEP-02.2.1 I Rev. 6 1 Page 267 of303 ATTACHMENT 2 Page 32 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. PC Radiation / RCS Activity Degradation Threat: Loss Threshold:

1. Drywell radiation

> 27 R/hr with reactor shutdown Definition(s): N/A BNP Basis: The Drywell High-Range Radiation Monitor (1 (2)D22-RI-4195, 1 (2)D22-RI-4196, 1 (2)D22-RI-4197, 1 (2)D22-RI-4198) reading of 27 R/hr is based on coolant activity at the Technical Specification limit of 4 pCi/gm 1-131).The containment radiation level of 27 R/hr is derived as follows: OPEP-03.6.3 Table 3 100% Cladding Damage column 'No Spray' for 1 hour after shutdown is 100,000 R/hr. Assuming that 300 pCi/gm 1-131 is approximately 2% cladding failure, a coolant activity of 4 pCi/gm 1-131 is ratioed to approximately 0.027% (0.00027) clad failure. Per Step 7.2.2.1, 0.00027 x 100,000 R/hr = 27 R/hr containment radiation corresponding to Technical Specification coolant activity. (ref. 1)The threshold value is only applicable with the reactor shutdown as the high range detectors normally read as high as 100 R/hr during power operations due to shine from the reactor.NEI 99-01 Basis: 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 D.1 since it indicates a loss of the RCS Barrier only.There is no Potential Loss threshold associated with Primary Containment Radiation. I OPEP-02.2.1 Rev. 6 Page 268 of 303 ATTACHMENT 2 Page 33 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases BNP Basis Reference(s):

1. OPEP-03.6.3 Estimate of the Extent of Core Damage Under Accident Conditions

2. NEI 99-01 Primary Containment Radiation RCS Loss 4.A I OPEP-02.2.1 I Rev. 6 1 Page 269 of 303 ATTACHMENT 2 Page 34 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

D. PC Radiation / RCS Activity Degradation Threat: Potential Loss Threshold: None IOPEP-02.2.1 I Rev. 6 Page 270 of 303 ATTACHMENT 2 Page 35 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. PC Integrity or Bypass Degradation Threat: Loss Threshold: None I OPEP-02.2.1 Rev. 6 Page 271 of 303 ' ATTACHMENT 2 Page 36 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. PC Integrity or Bypass Degradation Threat: Potential Loss Threshold: None I OPEP-02.2., I Rev. 6 1 Page 272 of303 ATTACHMENT 2 Page 37 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: F. SEC Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Site Emergency Coordinator that indicates loss of the RCS barrier Definition(s):

None BNP Basis: The Site Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences." Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to the recognition of the inability to reach safety acceptance criteria before completion of all checks." Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results." Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Site Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that may be used by the Site Emergency Coordinator in determining whether the RCS Barrier is lost.BNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A OPEP-02.2.1 Rev. 6 Page 273 of 303 ATTACHMENT 2 Page 38 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

F. SEC Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Site Emergency Coordinator that indicates potential loss of the RCS barrier Definition(s):

None BNP Basis: The Site Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to the inability to reach final safety acceptance criteria before completing all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Site Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that may be used by the Site Emergency Coordinator 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. BNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A 0PEP-02.2.1 I Rev. 6 Page 274 of 303 ATTACHMENT 2 Page 39 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

A. RPV Water Level Degradation Threat: Loss Threshold: None IOPEP-02.2.1 I Rev. 6 Page 275 of 303 ATTACHMENT 2 Page 40 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category: A. RPV Water Level Degradation Threat: Potential Loss Threshold:

1. Entry to SAMG-01 required Definition(s):

None BNP Basis: EOP-01-RVCP, EOP-01-LPC and EOP-01-RXFP specify the requirement for entry to SAMG-01 when core cooling is severely challenged. These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined. Entry to SAMG-01 is required when (ref. 1):* Reactor water level cannot be restored and maintained above -57.5 inches (Jet Pump Suction) with at least one core spray pump injecting into the reactor vessel* Reactor vessel water level cannot be restored and maintained above LL-4 (MSCRWL)* The reactor vessel flooding conditions cannot be restored and maintained (5 SRVs open and reactor vessel pressure more than 50 psig above suppression chamber pressure)" When at least 1 SRV cannot be opened and reactor vessel pressure cannot be restored and maintained above the minimum alternate reactor vessel flooding pressure (EOP-01-RXFP Table 1 values that are dependent on number of open SRVs)The above EOP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad.This threshold is also a Loss of the Fuel Clad barrier (FC Loss A.1). Since entry to SAMG-01 occurs after core uncovery has occurred a Loss of the RCS barrier exists (RCS Loss A.1).Entry to SAMG-01, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification. IOPEP-02.2.1 I Rev. 6 Page 276 of 303 ATTACHMENT 2 Page 41 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases NEI 99-01 Basis: The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold A.1.The Potential Loss requirement for entry to SAMG-01 indicates adequate core cooling cannot be restored and maintained and that core damage is possible. BWR EPGs/SAGs specify the conditions that require entry to SAMG-01. When entry to SAMG-01 is required, the EPGs are exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the inability to restore and maintain 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. BNP Basis Reference(s):

1. OSAMG-06.0 SAMG Primary Containment Flooding Basis Document 2. NEI 99-01 RPV Water Level PC Potential Loss 2.A I 0PEP-02.2.1 Rev. 6 Page 277 of 303 ATTACHMENT 2 Page 42 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

B. RCS Leak Rate Degradation Threat: Loss Threshold:

1. UNISOLABLE primary system leakage that results in exceeding EITHER of the following: " One or more Secondary Containment area radiation Maximum Safe Operating Limits (OEOP-03-SCCP Table 3)* One or more Secondary Containment area temperature Maximum Safe Operating Limits (OEOP-03-SCCP Table 1)Definition(s):

UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.BNP Basis: The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of unisolable primary system leakage outside the primary containment. The Maximum Safe Operating Limit values define this Containment barrier threshold because they are indicative of problems in the secondary containment that are spreading and pose a threat to achieving a safe plant shutdown. This threshold addresses problematic discharges outside primary containment that may not originate from a high-energy line break. The locations into which the primary system discharge is of concern correspond to the areas addressed in OEOP-03-SCCP, Secondary Containment Control Tables 3 and 1(ref.1) (see below). It should be noted that the Maximum Safe Radiation Operating Limits generally cannot be read in the Control Room and require local survey to assess.In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g.room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment. IOPEP-02.2.1 I Rev. 6 Page 278 of 303 ATTACHMENT 2 Page 43 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases NEI 99-01 Basis: 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 2.A this threshold would result in a Site Area Emergency. There is no Potential Loss threshold associated with Primary Containment Isolation Failure.BNP Basis Reference(s):

1. OEOP-03-SCCP Secondary Containment Control 2. NEI 99-01 RCS Leak Rate PC Loss 3.C IOPEP-02.2.1 I Rev. 6 1 Page 279 of 303 ATTACHMENT 2 Page 44 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

B. RCS Leak Rate Degradation Threat: Potential Loss Threshold: I None I OPEP-02.2.1 Rev. 6 1 Page 280 of 303 ATTACHMENT 2 Page 45 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Primary Containment Barrier: Category: C. PC Conditions Degradation Threat: Loss Threshold:

1. UNPLANNED rapid drop in Primary Containment pressure following Primary Containment pressure rise Definition(s):

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.BNP Basis: None NEI 99-01 Basis: 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. This threshold relies 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. BNP Basis Reference(s):

1. NEI 99-01 Primary Containment Conditions PC Loss 1.A IOPEP-02.2." I Rev. 6 Page 281 of 303 ATTACHMENT 2 Page 46 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment C. PC Conditions Category: Degradation Threat: Loss Threshold:

2. Primary Containment pressure response not consistent with LOCA conditions Definition(s):

None BNP Basis: The calculated pressure response of the containment is shown in Figure 6-11. Figure 6-11 shows that the maximum calculated drywell pressure is 48 psia (33 psig), which is well below the design allowable pressure of 62 psig (ref. 2). The primary containment pressure stabilizes at about 40 psia (25 psig), as shown on Figure 6-1.Due to conservatisms in LOCA analyses, actual pressure response is expected to be less than the analyzed response. For example, blowdown mass flowrate may be only 60-80% of the analyzed rate.NEI 99-01 Basis: 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. This threshold relies 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. IOPEP-02.2.1 I Rev. 6 Page 282 of 303 ATTACHMENT 2 Page 47 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases BNP Basis Reference(s):

1. BNP Updated FSAR Figure 6-11 2. BNP Updated FSAR section 6.2.1.1.1 3. NEI 99-01 Primary Containment Conditions PC Loss 1.B OPEP-02.2.1 Rev. 6 Page 283 of 303 ATTACHMENT 2 Page 48 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases CONTAINMENT PRESSURE RESPONSE TO DESIGN BASIS ACCIDENT LOCA AT 120 PERCENT UPRATE 50 40 3,A I 20 10 100 1000 10000 Time (seconds)1000D00 I OPEP-02.2.1 I Rev.6 6 Page 284 of303 ATTACHMENT 2 Page 49 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

B. PC Conditions Degradation Threat: Potential Loss Threshold:

1. Primary Containment pressure > 62 psig Definition(s):

None BNP Basis: When the primary containment exceeds the maximum allowable value (62 psig) (ref. 1), primary containment venting may be required even if offsite radioactivity release rate limits will be exceeded (ref. 2). The drywell and suppression chamber maximum allowable value of 62 psig is based on the primary containment design pressure as identified in the BNP accident analysis (ref. 1, 3). If this threshold is exceeded, a challenge to the containment structure has occurred because assumptions used in the accident analysis are no longer valid and an unanalyzed condition exists. This constitutes a Potential Loss of the Containment barrier even if a containment breach has not occurred.NEI 99-01 Basis: 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.BNP Basis Reference(s):

1. BNP Updated FSAR section 6.2.1.1.1 2. EOP-02-PCCP Primary Containment Control 3. BNP Updated FSAR section 6.2.1.1 4. NEI 99-01 Primary Containment Conditions PC Potential Loss I.A OPEP-02.2.

I Rev. 6 1 Page 285 of 303 ATTACHMENT 2 Page 50 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: B. PC Conditions Degradation Threat: Potential Loss Threshold:

2. Deflagration concentrations exist inside PC (H 2 > 6% AND 02 -- 5%)Definition(s):

None BNP Basis: Deflagration (explosive) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAMGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (ref. 1).Except for brief periods during plant startup and shutdown, oxygen concentration in the primary containment is maintained at insignificant levels by nitrogen inertion. The specified values for this Potential Loss threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, ref. 2) and readily recognizable because 6% hydrogen is well above the EOP-02-PCCP, Primary Containment Control, entry condition (ref. 2). The minimum global deflagration hydrogen/oxygen concentrations (6% and 5%, respectively) require intentional primary containment venting, which is defined to be a Loss of Containment (PC Loss C.4).Monitors CAC-AT-4409 and 4410 monitor hydrogen and oxygen concentration in the drywell during normal operation. These monitors are isolated by accident isolation signals. However, monitors CAC-AT-4409 and 4410 will be realigned to the primary containment for post-accident monitoring via an operator actuated isolation signal override circuit when directed by the EOPs.I OPEP-02.2.1 Rev. 6 1 Page 286 of 303 ý ATTACHMENT 2 Page 51 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases The oxygen and hydrogen concentrations from these two analyzers are recorded on two 4-channel recorders (CAC-AR-4409 and 4410) located on Panel XU-51. The indications are also displayed on the ERFIS. If concentrations exceed preset levels, recorder CAC-AR-4409 will annunciate the "Containment Atmosphere Division I 02 -H2 High" alarm in the Control Room and recorder CAC-AR-4410 will annunciate "Containment Atmosphere Division II 02 -H2 High" alarm. (ref. 3, 4).NEI 99-01 Basis: 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. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur.BNP Basis Reference(s):

1. BWROG EPG/SAG Revision 2, Sections PC/G 2. EOP-02-PCCP, Primary Containment Control 3. BNP Updated FSAR section 6.2.5.2.2 4. BNP System Description SD-04 Primary Containment

5. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.B OPEP-02.2.1 Rev. 6 Page 287 of 303 ATTACHMENT 2 Page 52 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Containment Barrier: Category: B. PC Conditions Degradation Threat: Potential Loss Threshold:

3. Heat Capacity Temperature Limit (HCTL) exceeded Definition(s):

None BNP Basis: This threshold is met when the final step of section SP/T in EOP-02-PCCP, Primary Containment Control, is reached (ref. 1, 2).NEI 99-01 Basis: The Heat Capacity Temperature Limit (HCTL) 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 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.I 0PEP-02.2.1 Rev. 6 Page 288 of 303 ATTACHMENT 2 Page 53 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases The HCTL 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. BNP Basis Reference(s):

1. EOP-01-NL EOP/SAMG Numerical Limits and Values 2. EOP-02-PCCP Primary Containment Control 3. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.C IOPEP-02.2.1 I Rev. 6 Page 289 of 303 ATTACHMENT 2 Page 54 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

D. PC Radiation / RCS Activity Degradation Threat: Loss Threshold: None I OPEP-02.2.1 Rev. 6 Page 290 of 303 ATTACHMENT 2 Page 55 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category: D. PC Radiation I RCS Activity Degradation Threat: Potential Loss Threshold:

1. Drywell radiation

> 20,000 R/hr Definition(s): None BNP Basis: The Drywell High-Range Radiation Monitor (1 (2)D22-RI-4195, 1 (2)D22-RI-4196, 1 (2)D22-RI-4197, 1(2)D22-RI-4198) reading of 20,000 R/hr indicates the release of reactor coolant into the drywell with elevated activity indicative of fuel damage.Based on 20% clad damage, a containment radiation level of 20,000 R/hr is derived as follows: OPEP-03.6.3 Table 3 100% Cladding Damage column 'No Spray' for 1 hour after shutdown is 100,000 R/hr. Per Step 7.2.2.1, 0.2 x 100,000 R/hr = 20,000 R/hr containment radiation corresponding to 20% clad damage.In order to reach this Containment barrier Potential Loss threshold, a loss of the RCS barrier (RCS Loss D.5) and a loss of the Fuel Clad barrier (FC Loss D.2) have already occurred. This threshold, therefore, represents at a General Emergency classification. NEI 99-01 Basis: 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. NUREG-1 228, 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. ATTACHMENT 2 OPEP-02.2.1 Rev. 6 T Page 291 of 303 Page 56 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases BNP Basis Reference(s):

1. OPEP-03.6.3 Estimate of the Extent of Core Damage Under Accident Conditions I OPEP-02.2.1 Rev. 6 Page 292 of 303 ATTACHMENT 2 Page 57 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

E. PC Integrity or Bypass Degradation Threat: Loss Threshold:

1. UNISOLABLE direct downstream pathway to the environment exists after Primary Containment isolation signal Definition(s):

UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally.BNP Basis: This threshold addresses failure of open isolation devices which should close upon receipt of a manual or automatic containment isolation signal resulting in a significant radiological release pathway directly to the environment. The concern is the unisolable open pathway to the environment. A failure of the ability to isolate any one line indicates a breach of primary containment integrity. As stated above, the adjective "Direct" modifies "release pathway" to discriminate against release paths through interfacing liquid systems. Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include unisolable Main steam line, HPCI steam line or RCIC steam line breaks, unisolable RWCU system breaks, and unisloable containment atmosphere vent paths. If the main condenser is available with an unisolable main steam line, there may be releases through the steam jet air ejectors and gland seal exhausters. These pathways are monitored, however, and do not meet the intent of a nonisolable release path to the environment. These minor releases are assessed using the Category R, Abnormal Rad Release / Rad Effluent, EALs.The existence of an in-line charcoal filter (SBGT) does not make a release path indirect since the filter is not effective at removing fission noble gases. Typical filters have an efficiency of 95-99% removal of iodine. Given the magnitude of the core inventory of iodine, significant releases could still occur. In addition, since the fission product release would be driven by boiling in the reactor vessel, the high humidity in the release stream can be expected to render the filters ineffective in a short period.I OPEP-02.2.1 Rev. 6 Page 293 of 303 1 ATTACHMENT 2 Page 58 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases The threshold is met if the breach is not isolable from the Control Room or an attempt for isolation from the Control Room has been made and was unsuccessful. An attempt for isolation from the Control Room should be made prior to the emergency classification. If operator actions from the Control Room are successful, this threshold is not applicable. Credit is not given for operator actions taken in-plant (outside the Control Room) to isolate the breach.EOP-02-PCCP, Primary Containment Control, Section PC/P may specify primary containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1). Under these conditions with a valid containment isolation signal, the Containment barrier should be considered lost.NEI 99-01 Basis: 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 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.BNP Basis Reference(s):

1. EOP-02-PCCP Primary Containment Control 2. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.A IOPEP-02.2.1 I Rev. 6 Page 294 of 303 ATTACHMENT 2 Page 59 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

E. PC Integrity or Bypass Degradation Threat: Loss Threshold:

2. Intentional Primary Containment venting per EOPs Definition(s):

None BNP Basis: EOP-02-PCCP, Primary Containment Control, may specify primary containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1, 2). The threshold is met when the operator begins venting the primary containment in accordance with EOP-01 -SEP-01, not when actions are taken to bypass interlocks prior to opening the vent valves. Purge and vent actions specified in step PC/P-03 to control drywell pressure below the drywell high pressure scram setpoint or in section PC/H does not meet this threshold because such action is only permitted if offsite radioactivity release rates will remain below the ODCM limits.NEI 99-01 Basis: 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 control to the secondary containment and/or the environment is a Loss of the Containment. 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. BNP Basis Reference(s):

1. EOP-02-PCCP Primary Containment Control 2. EOP-01-SEP-01 Primary Containment Venting 3. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.B OPEP-02.2.1 Rev. 6 Page 295 of 303 ATTACHMENT 2 Page 60 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

E. PC Integrity or Bypass Degradation Threat: Potential Loss Threshold: None OPEP-02.2.1 Rev. 6 Page 296 of 303 ATTACHMENT 2 Page 61 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category: F. SEC Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Site Emergency Coordinator that indicates loss of the Primary Containment barrier Definition(s):

None BNP Basis: The Site Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks." Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.* Dominant accident seguences lead to degradation of all fission product barriers and likely entry to the EOPs. The Site Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that may be used by the Site Emergency Coordinator in determining whether the Primary Containment Barrier is lost.BNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A IOPEP-02.2.1 I Rev. 6 1 Page 297 of 303 ATTACHMENT 2 Page 62 of 62 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Primary Containment Category:

F. SEC Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Site Emergency Coordinator that indicates potential loss of the Primary Containment barrier Definition(s):

None BNP Basis: The Site Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks." Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Site Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. NEI 99-01 Basis: This threshold addresses any other factors that may be used by the Site Emergency Coordinator in determining whether the Primary Containment Barrier is lost.BNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A OPEP-02.2.1 Rev. 6 Page 298 of 303 ATTACHMENT 3 Page 1 of 4 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases Background NEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located.These areas are intended to be plant operating mode dependent. Specifically the Developers Notes For AA3 and HA5 states: The "site-specific list of plant rooms or areas with entry-related mode applicability identified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area.The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections). Further, as specified in IC HA5: The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas.Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope.I OPEP-02.2.1 Rev. 6 Page 299 of 303 ATTACHMENT 3 Page 2 of 4 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases BNP Table R-2 and H-2 Bases A review of station operating procedures identified the following mode dependent in-plant actions and associated safe shutdown areas that are required for normal plant operation, cooldown or shutdown: Location-Modes- Modes-Safe Shutdown Area 1,2 3,4,5-17 North RHR Unit-1 RHR Equipment. RHR Shut Down Cooling (SDC)-No entry required RHR Pump Discharge Isolation Valves Ell-F018 A&C Inventory Control Equipment-No entry required Reactivity Control.-No entry required-17 North RHR Unit-2 RHR Equipment. RHR Shut Down Cooling (SDC)-No entry required RHR Pump Discharge Isolation Valves Ell-F018 A&C Inventory Control Equipment. -No entry required Reactivity Control.-No entry required-17 South RHR Unit-1 RHR Equipment. RHR Shut Down Cooling (SDC)-No entry required RHR Pump Discharge Isolation Valves Ell-F018 B&D Inventory Control Equipment. -No entry required Reactivity Control.-No entry required-17 South RHR Unit-2 RHR Equipment. RHR Shut Down Cooling (SDC)-No entry required RHR Pump Discharge Isolation Valves Ell-F018 B&D Inventory Control Equipment. -No entry required Reactivity Control.-No entry required-17 North Core Spray Core Spray Equipment Inventory Control-No entry required -No entry required-17 South Core Spray Core Spray Equipment Inventory Control.-No entry required -No entry required Service Water Building 20' Heat Sink equipment. Heat Sink equipment. -No entry required -No entry required I 0PEP-02.2.1 Rev. 6 1 Page 300 of 303 ATTACHMENT 3 Page 3 of 4 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases Control Building HVAC Room Habitability. Habitability.(Turbine Building 70') -No entry required -No entry required Reactor Building HVAC Habitability. Habitability.(Reactor Building 80' West) -No entry required -No entry required Emergency Diesel Generators Electrical Power, Local control Electrical Power, Local control (EDG Building 20') parameters at EDG panel. parameters at EDG panel.-No entry required -No entry required Electrical Power. Electrical Power.Emergency Diesel Generators 4-Day Tank Rooms -No entry required -No entry required EDG Building HVAC (EDG Habitability. Habitability. Building 70') -No entry required -No entry required Electrical Power. Electrical Power.1 &2) -No entry required -No entry required 4160 VAC (EDG building 70') -No entry required -No entry required 480 VAC (EDG Building 20')N -No entry required -No entry required& S ends 120 VAC Vital (Cable Spread U- -No entry required -No entry required 1 & U-2)Train A & B DC (Battery Rooms -No entry required -No entry required U-1 & U-2)Reactor Building 20' East & -No entry required RHR SDC.West MCC Areas -1 (2) Ell -F009 & F008 valve breakers (RHR SDC Suction isolation valves)-1 (2) ElI -F006 A-D valve breakers, (RHR pump suction valves)-RHR SDC Suction Fill &Vent valves (Manual Valves)-RHR suction pipe flush valve breakers (ElI-FO11A & B, Ell-V33, Ell-V32 Reactor Building 20' Pipe No entry required RHR SDC Tunnel RHR SDC Suction Fill &Vent valves (Manual Valves)OPEP-02.2.1 Rev. 6 Page 301 of 303 ATTACHMENT 3 Page 4 of 4 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases Table R-2 & H-2 Results Table R-2 & H-2 Safe Operation & Shutdown Areas Room/Area Mode Applicability Reactor Building -17' North RHR Unit-1 & 2 3, 4, 5 Reactor Building -17' South RHR Unit-1 & 2 3, 4, 5 Reactor Building 20' East & West MCC Areas Unit-1 & 2 3, 4, 5 Reactor Building 20' Pipe Tunnel Unit-1 & 2 3, 4, 5 Plant Operating Procedures Reviewed 1. Unit 1 & 2 RHR OP-17" Shutdown Cooling" Low Pressure Coolant Injection 2.3.4.5.6.7.Unit-1 & 2 UAT Backfeed OP-50 EDG Operation OP-50.1, OP-39 Unit-1 & 2 Service Water OP-43 Unit-1 & 2 Core Spray Control Building Ventilation System 20P-37 Defense In Depth AP-22 OPEP-02.2.1 Rev. 6 Page 302 of 303 REVISION

SUMMARY

Revision 6 of OPEP-02.2.1 consists of the following changes:[xxxx]OPEP-02.2.1 Rev. 6 Page 303 of 303 BSEP 15-0010 Enclosure 4 Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 License Amendment Request to Adopt Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors" BSEP EAL Comparison Matrix F DUKE ENERGYBrunswick Nuclear Plant NEI 99-01 Revision 6 EAL Comparison Matrix Revision 0 [Draft D4 8/16/14] EAL Comparison Matrix OSSI Project #13-0703 BNP Table of Contents Section Paqe Introduction

1 Comparison Matrix Format ------------------------------------------------------------------------ 1 EAL Wording ------------------------------------------------------------------------------------------------------------------------------------------------------ 1 EAL Emphasis Techniques

1 Global Differences

2 Differences and Deviations

3 Category A -Abnormal Rad Levels / Rad Effluent ------------------------------------------------------ 12 Category C -Cold Shutdown / Refueling System Malfunction

29 Category D -Permanently Defueled Station Malfunction

48 Category E -Independent Spent Fuel Storage Installation (ISFSI) ------------------------------------------- 50 Category F -Fission Product Barrier Degradation

52 Category H -Hazards and Other Conditions Affecting Plant Safety ------------------------------------------ 64 Category S -System Malfunction

83 Table 1 -BNP 2 EAL Categories/Subcategories

5 Table 2 -NEI / BNP EAL Identification Cross-Reference

6 Table 3 -Summary of Deviations

11 iofi EAL Comparison Matrix OSSI Project #13-0703 BNP Introduction This document provides a line-by-line comparison of the Initiating Conditions (ICs), Mode Applicability and Emergency Action Levels (EALs) in NEI 99-01 Rev. 6 Final, Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML1 10240324, and the Brunswick Nuclear Plant (BNP) ICs, Mode Applicability and EALs. This document provides a means of assessing BNP differences and deviations from the NRC endorsed guidance given in NEI 99-01. Discussion of BNP EAL bases and lists of source document references are given in the EAL Technical Bases Document. It is, therefore, advisable to reference the EAL Technical Bases Document for background information while using this document.Comparison Matrix Format The ICs and EALs discussed in this document are grouped according to NEI 99-01 Recognition Categories. Within each Recognition Category, the ICs and EALs are listed in tabular format according to the order in which they are given in NEI 99-01. Generally, each row of the comparison matrix provides the following information:

NEI EAL/IC identifier
NEI EAL/IC wording* BNP EAL/IC identifier
BNP EAL/IC wording* Description of any differences or deviations EAL Wording In Section 4.1, NEI recommends the following: "The guidance in NEI 99-01 is not intended to be applied to plants "as-is"; however, developers should attempt to keep their site-specific schemes as close to the generic guidance as possible.

The goal is to meet the intent of the generic Initiating Conditions (ICs) and Emergency Action Levels (EALs) within the context of site-specific characteristics -locale, plant design, operating features, terminology, etc.Meeting this goal will result in a shorter and less cumbersome NRC review and approval process, closer alignment with the schemes of other nuclear power plant sites and better positioning to adopt future industry-wide scheme enhancements" To assist the Site Emergency Coordinator (SEC), the BNP EALs have been written in a clear and concise style (to the extent that the differences from the NEI EAL wording could be reasonably documented and justified). As a result, any unnecessary words have been removed from the BNP EALs to reduce EAL-user reading burden to the extent practicable. The wording reduction gained from elimination of a few characters in a given EAL may not appear to be advantageous within the context of one EAL.When applied to the composite set of EALs, however, significant gains are realized and reading efficiency is improved. This supports timely and accurate classification in the tense atmosphere of an emergency event. The EAL differences introduced to reduce reading burden comprise almost all of the differences justified in this document.EAL Emphasis Techniques Due to the width of the table columns and table formatting constraints in this document, line breaks and indentation may differ slightly from the appearance of comparable wording in the source documents. NEI 99-01 is the source document for the NEI EALs; the BNP EAL Technical Bases Document for the BNP EALs.Development of the BNP IC/EAL wording has attempted to minimize inconsistencies and apply sound human factors principles. As a result, differences occur between NEI and BNP ICs/EALs for these reasons alone.When such difference may infer a technical difference in the associated NEI IC/EAL, the difference is identified and a justification provided.The print and paragraph formatting conventions summarized below guide presentation of the BNP EALs in accordance with the EAL writing criteria.Space restrictions in the EAL table of this document sometimes override these criteria in cases when following the criteria would introduce undesirable complications in the EAL layout." Upper case-bold print is used for the logic terms AND, OR and EITHER.* Bold font is used for certain logic terms, negative terms (not, cannot, etc.), any, all.* Upper case print is reserved for defined terms, acronyms, system abbreviations, logic terms (and, or, etc. when not used as a conjunction), annunciator window engravings. 1 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP" Three or more items in a list are normally introduced with "Any of the following..." or "All of the following..." Items of the list begin with bullets when a priority or sequence is not inferred.* The use of AND/OR logic within the same EAL has been avoided when possible. When such logic cannot be avoided, indentation and separation of subordinate contingent phrases is employed.Global Differences The differences listed below generally apply throughout the set of EALs and are not repeated in the Justification sections of this document. The global differences do not decrease the effectiveness of the intent of NEI 99-01.1. The NEI phrase "Notification of Unusual Event" has been changed to"Unusual Event" or abbreviated "UE" to reduce EAL-user reading burden.2. The NEI term Emergency Director had been replaced with the BNP-specific title Site Emergency Coordinator (SEC).3. NEI 99-01 IC Example EALs are implemented in separate plant EALs to improve clarity and readability. For example, NEI lists all IC HU3 Example EALs under one IC. The corresponding BNP EALs appear as unique EALs (e.g., HU3.1 through HU3.4).4. Mode applicability identifiers (numbers/letter) modify the NEI 99-01 mode applicability names as follows: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown, 4 -Cold Shutdown, 5 -Refuel, D -Defueled, and All. NEI 99-01 defines Defueled as follows: "Reactor Vessel contains no irradiated fuel (full core off-load during refueling or extended outage)." 5. NEI 99-01 uses the terms greater than, less than, greater than or equal to, etc. in the wording of some example EALs. For consistency and reduce EAL-user reading burden, BNP has adopted use of Boolean symbols in place of the NEI 99-01 text modifiers within the EAL wording.6. "min." is the standard abbreviation for "minutes" and is used to reduce EAL user reading burden.7. IC/EAL identification: " NEI Recognition Category A "Abnormal Radiation Levels/Radiological Effluents" has been changed to Category R"Abnormal Rad Levels / Rad Effluent." The designator "R" is more intuitively associated with radiation (rad) or radiological events. NEI IC designators beginning with "A" have likewise been changed to "R."" NEI 99-01 defines the thresholds requiring emergency classification (example EALs) and assigns them to ICs which, in turn, are grouped in "Recognition Categories." The Recognition Categories, however, are so broad and the IC descriptions are so varied that an EAL is difficult to locate in a timely manner when the EAL-user must refer to a set of EALs with the NEI organization and identification scheme. The NEI document clearly states that the EAL/IC/Recognition Category scheme is not intended to be the plant-specific EAL scheme for any plant, and appropriate human factors principles should be applied to development of an EAL scheme that helps the EAL-user make timely and accurate classifications. BNP endeavors to optimize the NEI EAL organization and identification scheme to enhance usability of the plant-specific EAL set. To this end, the BNP IC/EAL scheme includes the following features: a. Division of the NEI EAL set into three groups: o EALs applicable under all plant operating modes -This group would be reviewed by the EAL-user any time emergency classification is considered. o EALs applicable only under hot operating modes -This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup or Power Operation mode.o EALs applicable only under cold operating modes -This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refuel or Defueled mode.The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the 2 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.

b. Within each of the above three groups, assignment of EALs to categories/subcategories

-Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user.Subcategories are used as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The BNP EAL categories/subcategories and their relationship to NEI Recognition Categories are listed in Table 1.c. Unique identification of each EAL -Four characters comprise the EAL identifier as illustrated in Figure 1.Figure 1 -EAL Identifier EAL Identifier XXX.X Category (R, H. E, S. F. C) -L Sequential number within subcategory/classification Emergency classification(.S. A.U Subcategory number (1 if no subicategory) The first character is a letter associated with the category in which the EAL is located. The second character is a letter associated with the emergency classification level (G for General Emergency, S for Site Area Emergency, A for Alert, and U for Notification of Unusual Event). The third character is a number associated with one or more subcategories within a given category. Subcategories are sequentially numbered beginning with the number "1". If a category does not have a subcategory, this character is assigned the number "1 ". The fourth character is a number preceded by a period for each EAL within a subcategory. EALs are sequentially numbered within the emergency classification level of a subcategory beginning with the number "1 ".The EAL identifier is designed to fulfill the following objectives: o Uniqueness -The EAL identifier ensures that there can be no confusion over which EAL is driving the need for emergency classification. o Speed in locating the EAL of concern -When the EALs are displayed in a matrix format, knowledge of the EAL identifier alone can lead the EAL-user to the location of the EAL within the classification matrix. The identifier conveys the category, subcategory and classification level. This assists ERO responders (who may not be in the same facility as the SEC) to find the EAL of concern in a timely manner without the need for a word description of the classification threshold. o Possible classification upgrade -The category/subcategory/identifier scheme helps the EAL-user find higher emergency classification EALs that may become active if plant conditions worsen.Table 2 lists the BNP ICs and EALs that correspond to the NEI ICs/Example EALs when the above EAL/IC organization and identification scheme is implemented. Differences and Deviations In accordance NRC Regulatory Issue Summary (RIS) 2003-18 "Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels" Supplements 1 and 2, a difference is an EAL change in which 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 BNP EAL. A deviation is an EAL change in which 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 BNP proposed EAL.Administrative changes that do not actually change the textual content are neither differences nor deviations. Likewise, any format change that does not alter the wording of the IC or EAL is considered neither a difference nor a deviation. 3 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP The following are examples of differences: " Choosing the applicable EAL based upon plant type (i.e., BWR vs.PWR).* Using a numbering scheme other than that provided in NEI 99-01 that does not change the intent of the overall scheme.* Where the NEI 99-01 guidance specifically provides an option to not include an EAL if equipment for the EAL does not exist at BNP (e.g., automatic real-time dose assessment capability)." Pulling information from the bases section up to the actual EAL that does not change the intent of the EAL." Choosing to state ALL Operating Modes are applicable instead of stating N/A, or listing each mode individually under the Abnormal Rad Level/Radiological Effluent and Hazard and Other Conditions Affecting Plant Safety sections.* Using synonymous wording (e.g., greater than or equal to vs. at or above, less than or equal vs. at or below, greater than or less than vs. above or below, etc.)* Adding BNP equipment/instrument identification and/or noun names to EALs.* Combining like ICs that are exactly the same but have different operating modes as long as the intent of each IC is maintained and the overall progression of the EAL scheme is not affected." Any change to the IC and/or EAL, and/or basis wording, as stated in NEI 99-01, that does not alter the intent of the IC and/or EAL, i.e., the IC and/or EAL continues to: o Classify at the correct classification level.o Logically integrate with other EALs in the EAL scheme.o Ensure that the resulting EAL scheme is complete (i.e., classifies all potential emergency conditions). The following are examples of deviations:

Use of altered mode applicability.
Altering key words or time limits.* Changing words of physical reference (protected area, safety-related equipment, etc.)." Eliminating an IC. This includes the removal of an IC from the Fission Product Barrier Degradation category as this impacts the logic of Fission Product Barrier ICs.* Changing a Fission Product Barrier from a Loss to a Potential Loss or vice-versa.
Not using NEI 99-01 definitions as the intent is for all NEI 99-01 users to have a standard set of defined terms as defined in NEI 99-01.Differences due to plant types are permissible (BWR or PWR).Verbatim compliance to the wording in NEI 99-01 is not necessary as long as the intent of the defined word is maintained.

Use of the wording provided in NEI 99-01 is encouraged since the intent is for all users to have a standard set of defined terms as defined in NEI 99-01." Any change to the IC and/or EAL, and/or basis wording as stated in NEI 99-01 that does alter the intent of the IC and/or EAL, i.e., the IC and/or EAL: o Does not classify at the classification level consistent with NEI 99-01.o Is not logically integrated with other EALs in the EAL scheme.o Results in an incomplete EAL scheme (i.e., does not classify all potential emergency conditions). The "Difference/Deviation Justification" columns in the remaining sections of this document identify each difference between the NEI 99-01 IC/EAL wording and the BNP IC/EAL wording. An explanation that justifies the reason for each difference is then provided. If the difference is determined to be a deviation, a statement is made to that affect and explanation is given that states why classification may be different from the NEI 99-01 IC/EAL and the reason for its acceptability. In all cases, however, the differences and deviations do not decrease the effectiveness of the intent of NEI 99-01. A summary list of BNP EAL deviations from NEI 99-01 is given in Table 3.4 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Table 1 -BNP EAL Categories/Subcategories BNP EALs NEI Category Subcategory Recognition Category Group: Any Operating Mode: 1 -Radiological Effluent Abnormal Rad Levels/Radiological Effluent R -Abnormal Rad Levels/Rad Effluent 2 -Irradiated Fuel Event ICs/EALs 3 -Area Radiation Levels 1 -Security Hazards and Other Conditions Affecting 2 -Seismic Event Plant Safety ICs/EALs 3 -Natural or Technological Hazard H -Hazards and Other Conditions Affecting 4 -Fire Plant Safety 5 -Hazardous Gas 6 -Control Room Evacuation 7 -SEC Judgment E -ISFSI 1 -Confinement Boundary ISFSI ICs/EALs Group: Hot Conditions: 1 -Loss of Emergency AC Power System Malfunction ICs/EALs 2 -Loss of Vital DC Power 3 -Loss of Control Room Indications 4 -RCS Activity S -System Malfunction 5 -RCS Leakage 6 -RPS Failure 7 -Loss of Communications 8 -Hazardous Event Affecting Safety Systems F -Fission Product Barrier Degradation None Fission Product Barrier ICs/EALs Group: Cold Conditions: 1 -RCS Level Cold Shutdown./ Refueling System 2 -Loss of Emergency AC Power Malfunction ICs/EALs C -Cold Shutdown/Refueling System 3 -RCS Temperature Malfunction 4 -Loss of Vital DC Power 5 -Loss of Communications 6 -Hazardous Event Affecting Safety Systems 5 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Table 2 -NEI / BNP EAL Identification Cross-Reference NEI BNP Example Category and Subcategory EAL AC 1EAL AU1 1 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RU1.1 AU1 2 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RU1.1 AU2 3 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RU1.2 AU2 1 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RU2.1 AA1 1 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RA1.1 AA1 2 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RA1.2 AA1 3 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RA1.3 AA1 4 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RA1.4 AA2 1 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RA2.1 AA2 2 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RA2.2 AA2 3 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RA2.3 AA3 1 R -Abnormal Rad Levels / Rad Effluent, 3 -Area Radiation Levels RA3.1 AA3 2 R -Abnormal Rad Levels / Rad Effluent, 3 -Area Radiation Levels RA3.2 AS1 1 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RS1.1 AS1 2 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RS1.2 AS1 3 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RS1.3 6 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP NEI BNP Example Category and Subcategory EAL A2 1EAL AS2 1 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RS2.1 AG1 1 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RG1.1 AG1 2 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RG1.2 AG1 3 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RG1.3 AG2 1 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RG2.1 CUl 1 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CU1.1 CU1 2 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CU1.2 CU2 1 C -Cold SD/ Refueling System Malfunction, 2 -Loss of Emergencyl AC Power CU2.1 CU3 1 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CU3.1 CU3 2 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CU3.2 CU4 1 C -Cold SD/ Refueling System Malfunction, 4 -Loss of Vital DC Power CU4.1 CU5 1,2, 3 C -Cold SD/ Refueling System Malfunction, 5 -Loss of Communications CU5.1 CA1 1 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CA1.1 CA1 2 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CA1.2 CA2 1 C -Cold SD/ Refueling System Malfunction, 1 -Loss of Emergency AC Power CA2.1 CA3 1 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CA3.1 CA3 2 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CA3.1 CA6 1 C -Cold SD/ Refueling System Malfunction, 6 -Hazardous Event Affecting Safety Systems CA6.1 7 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP NEI BNP IC Example Category and Subcategory EAL EAL CS1 1 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CS1.1 CS1 2 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CS1.2 CS1 3 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CS1.3 CG1 1 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CS1.1 CG1 2 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CG1.2 E-HU1 1 E -ISFSI EU1.1 FA1 1 F -Fission Product Barrier Degradation FA1.1 FS1 1 F -Fission Product Barrier Degradation FS1.1 FG1 1 F -Fission Product Barrier Degradation FG1.1 HU1 1 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HU1.1 HU1 2 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HU1.2 HU1 3 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HU1.3 HU2 1 H -Hazards and Other Conditions Affecting Plant Safety, 2 -Seismic Event HU2.1 HU3 1 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.1 HU3 2 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.2 HU3 3 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.3 HU3 4 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.4 HU3 5 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.5 8 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP NEI BNP IC Example Category and Subcategory EAL EAL HU4 1 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire HU4.1 HU4 2 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire HU4.2 HU4 3 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire HU4.3 HU4 4 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire HU4.4 HU7 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -SEC Judgment HU7.1 HAl 1 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HAl .1 HAl 2 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HA1.2 HA5 1 H -Hazards and Other Conditions Affecting Plant Safety, 5 -Hazardous Gas HA5.1 HA6 1 H -Hazards and Other Conditions Affecting Plant Safety, 6 -Control Room Evacuation HA6.1 HA7 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -SEC Judgment HA7.1 HS1 1 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HS1.1 HS6 1 H -Hazards and Other Conditions Affecting Plant Safety, 6 -Control Room Evacuation HS6.1 HS7 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -SEC Judgment HS7.1 HG1 1 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HG1.1 HG7 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -SEC Judgment HG7.1 SUl 1 S -System Malfunction, 1 -Loss of Emergency AC Power SU1.1 SU2 1 S -System Malfunction, 3 -Loss of Control Room Indications SU3.1 SU3 1 S -System Malfunction, 4 -RCS Activity SU4.1 9 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP NEI BNP Example Category and Subcategory EAL EAL SU3 2 S -System Malfunction, 4 -RCS Activity SU4.2 SU4 1,2, 3 S -System Malfunction, 5 -RCS Leakage SU5.1 SU5 1 S -System Malfunction, 6 -RPS Failure SU6.1 SU5 2 S -System Malfunction, 6 -RPS Failure SU6.2 SU6 1,2,3 S -System Malfunction, 7 -Loss of Communications SU7.1 SU7 1,2 N/A (PWR only) N/A SA1 1 S -System Malfunction, 1 -Loss of Emergency AC Power SA1.1 SA2 1 S -System Malfunction, 3 -Loss of Control Room Indications SA3.1 SA5 1 S -System Malfunction, 6 -RPS Failure SA6.1 SA9 1 S -System Malfunction, 8 -Hazardous Event Affecting Safety Systems SA8.1 SS1 1 S -System Malfunction, 1 -Loss of Essential AC Power SS1.1 SS5 1 S -System Malfunction, 6 -RPS Failure SS6.1 SS8 1 S -System Malfunction, 2 -Loss of Vital DC Power SS2.1 SG1 1 S -System Malfunction, 1 -Loss of Emergency AC Power SG1.1 SG8 2 S -System Malfunction, 1 -Loss of EmergencyAC Power SG1.2 10 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Table 3 -Summary of Deviations NEI BNP IC Example EAL EAL Description None None N/A N/A 11 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Category A Abnormal Rad Levels/ Radiological Effluent 12 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP NEI IC# NEI IC Wording and Mode BNP BNP IC Wording and Mode Difference/Deviation Justification Applicability IC#(s) Applicability AUL1 Release of gaseous or liquid RU1 Release of gaseous or liquid The BNP ODCM is the site-specific effluent release radioactivity greater than 2 times radioactivity greater than 2 times the controlling document.the (site-specific effluent release ODCM limits for 60 minutes or longer.controlling document) limits for MODE: All 60 minutes or longer.MODE: All NEI Ex. NEI Example EAL Wording BNP BNP EAL Wording Difference/Deviation Justification EAL # EAL # N A odn Reading on ANY effluent Reading on any Table R-1 effluent Example EALs #1 and #2 have been combined into a single radiation monitor greater than 2 radiation monitor > column "UE" for > EAL to simplify presentation. times the (site-specific effluent 60 min. The NEI phrase "...effluent radiation monitor greater than 2 release controlling document) (Notes 1, 2, 3) times the (site-specific effluent release controlling limits for 60 minutes or longer: document)" and "effluent radiation monitor greater than 2 (site-specific monitor list and times the alarm setpoint established by a current radioactivity threshold values corresponding discharge permit" have been replaced with "...any Table R-1 to 2 times the controlling RU1.1 effluent radiation monitor > column "UE".document limits) UE thresholds for all BNP continuously monitored gaseous 2 Reading on ANY effluent and liquid release pathways are listed in Table R-1 to radiation monitor greater than 2 consolidate the information in a single location and, thereby, times the alarm setpoint simplify identification of the thresholds by the EAL user. The established by a current values shown in Table R-1 column "UE", consistent with the radioactivity discharge permit for NEI bases, represent two times the ODCM release limits for 60 minutes or longer, both liquid and gaseous release.3 Sample analysis for a gaseous or RU1.2 Sample analysis for a gaseous or The BNP ODCM is the site-specific effluent release liquid release indicates a liquid release indicates a concentration controlling document.concentration or release rate or release rate > 2 x ODCM limits for >greater than 2 times the (site- 60 min.specific effluent release (Notes 1, 2)controlling document) limits for 13 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP NEI Ex. NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 60 minutes or longer.Notes

The Emergency Director N/A Note 1: The SEC should declare the The classification timeliness note has been standardized should declare the Unusual event promptly upon across the BNP EAL scheme by referencing the "time limit" Event promptly upon determining that time limit specified within the EAL wording.determining that 60 minutes has been exceeded, or will has been exceeded, or will likely be exceeded.likely be exceeded.

Note 2: If an ongoing release is* If an ongoing release is detected and the release The classification timeliness note has been standardized detected and the release start time is unknown, across the BNP EAL scheme by referencing the "time limit" start time is unknown, assume that the release specified within the EAL wording.assume that the release duration has exceeded the duration has exceeded 60 specified time limit.minutes. Note 3: If the effluent flow past an* If the effluent flow past an effluent monitor is known to None effluent monitor is known to have stopped, indicating that have stopped due to actions the release path is isolated, to isolate the release path, the effluent monitor reading then the effluent monitor is no longer VALID for reading is no longer valid for classification purposes.classification purposes.14 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Main Stack Rad D12-RM-23S 2.13E+09 pCi/sec 2.13E+08 pCi/sec 2.13E+07 VCi/sec 1.80E+6 p.Ci/secReactor Bldg Vent Noble CAC-AQH-1264-3

---- ---- 6.14E+4 cpm o Gas Turbine Bldg Vent D12-RM-23 1.07E+08 pCi/sec 1.07E+07 pCi/sec 1.07E+06 pCi/sec 1.13E+4 pCi/sec Service Water Effluent D12-RM-K605

---- 2 x hi alarm Rad 27-J Radwaste Effluent Rad D12-RM-K604

............- 2 x hi-hi alarm 15 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP NEI IC# NEI IC Wording and Mode BNP BNP IC Wording and Mode Difference/Deviation Justification Applicability IC#(s) Applicability AU2 UNPLANNED loss of water level RU2 Unplanned loss of water level above None above irradiated fuel. irradiated fuel MODE: All MODE: All NEI Ex. BNP EA NEI Example EAL Wording EAL BNP EAL Wording Difference/Deviation Justification EAL # EAL #a. UNPLANNED water level RU2.1 UNPLANNED water level drop in the Site-specific level indications and area radiation monitors are drop in the REFUELING REFUELING PATHWAY as indicated listed in bullet format for clarification. PATHWAY as indicated by by low water level alarm (A-04 6-6) or ANY of the following: indication (site-specific level indications). AND AND UNPLANNED rise in area radiation UNPLANNEDAiseDinrare radiatio b. UNPLANNED rise in area levels as indicated by any of the radiation levels as indicated following radiation monitors: by ANY of the following radiation monitors.

ARM Channel 26 New Fuel Vault (site-specific list of area radiation monitors)

P ARM Channel 27 North of Fuel Pool 0 ARM Channel 28 Between Reactor and Fuel Pool* ARM Channel 29 Cask Wash Area 16 of 107 EAL Comparison Matrix OSS, Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification AA1 Release of gaseous or liquid RA1 Release of gaseous or liquid None radioactivity resulting in offsite radioactivity resulting in offsite dose dose greater than 10 mrem TEDE greater than 10 mrem TEDE or 50 or 50 mrem thyroid CDE. mrem thyroid CDE MODE: All MODE: All NEI Ex. NEI Example EAL Wording BNP BNP EAL Wording Difference/Deviation Justification EAL # EAL #Reading on ANY of the following RA1.1 In the absence of real-time dose The BNP radiation monitors that detect radioactivity effluent radiation monitors greater than assessment, reading on any Table R-1 release to the environment are listed in Table R-1. UE, Alert, the reading shown for 15 minutes effluent radiation monitor > column SAE and GE thresholds for all BNP continuously monitored or longer: "ALERT" for > 15 min. (Notes 1, 2, 3, gaseous release pathways are listed in Table R-1 to i4) consolidate the information in a single location and, thereby, (site-specific monitor list and simplify identification of the thresholds by the EAL-user.threshold values) Added "In the absence of real-time dose assessment" to re-enforce Note 4.2 Dose assessment using actual RA1.2 Dose assessment using actual The site boundary is the site-specific receptor point.meteorology indicates doses meteorology indicates doses > 10 greater than 10 mrem TEDE or mrem TEDE or 50 mrem thyroid CDE 50 mrem thyroid CDE at or at or beyond the SITE BOUNDARY beyond (site-specific dose (Notes 3, 4)receptor point).3 Analysis of a liquid effluent RA1.3 Analysis of a liquid effluent sample The site boundary is the site-specific receptor point.sample indicates a concentration indicates a concentration or release or release rate that would result rate that would result in doses > 10 in doses greater than 10 mrem mrem TEDE or 50 mrem thyroid CDE TEDE or 50 mrem thyroid CDE at at or beyond the SITE BOUNDARY for or beyond (site-specific dose 60 min. of exposure (Notes 1, 2)receptor point) for one hour of exposure.17 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP 4 Field survey results indicate EITHER of the following at or beyond (site-specific dose receptor point): " 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. RA1.4 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: " Closed window dose rates > 10 mR/hr expected to continue for> 60 min." Analyses of field survey samples indicate thyroid CDE >50 mrem for 60 min. of inhalation.(Notes 1,2)The site boundary is the site-specific receptor point.Notes* The Emergency Director should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes." The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results N/A Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the The classification timeliness note has been standardized across the BNP EAL scheme by referencing the "time limit" specified within the EAL wording.The classification timeliness note has been standardized across the BNP EAL scheme by referencing the "time limit" specified within the EAL wording.None Incorporated site-specific EAL numbers associated with generic EAL#1.18 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP from a dose assessment results from a dose using actual meteorology are assessment using actual available. meteorology are available. 19 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification Significant lowering of water RA2 Significant lowering of water level above None level above, or damage to, or damage to irradiated fuel irradiated fuel. MODE: All MODE: All NEI Ex. BNP EA E NEI Example EAL Wording EAL BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 Uncovery of irradiated fuel in the RA2.1 Uncovery of irradiated fuel in the None REFUELING PATHWAY. REFUELING PATHWAY 2 Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of the following radiation monitors: (site-specific listing of radiation monitors, and the associated readings, setpoints and/or alarms)RA2.2 Damage to irradiated fuel resulting in a release of radioactivity AND Any of the following radiation monitor indications: " Reactor Bldg Vent Rad Monitor Channel A or B (> 3 mR/hr)" ARM Channel 26 New Fuel Vault (> 6 mR/hr)" ARM Channel 27 North of Fuel Pool (>10 mR/hr)" ARM Channel 28 Between Reactor and Fuel Pool (> 1000 mR/hr)" ARM Channel 29 Cask Wash Area (>40 mR/hr)Deleted the NEI phrase "from the fuel" because it is redundant to the preceding phrase "irradiated fuel." Site-specific list of radiation monitors are listed in bullet format for clarification. Listed in bullet format for clarification. +Lowering of spent fuel pool level to (site-specific Level 2 value).[See Developer Notes]RA2.3 Lowering of spent fuel pool level to< 105 ft. 3 in. ele.Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1 -116 ft. 1 in. ele.), SFP level 10 ft. above the top of the fuel racks (Level 2 -105 ft. 3 in. ele.) and SFP level at the 20 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Itop of the fuel racks (Level 3 -95 ft. 3 in. ele.).BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification AA3 Radiation levels that impede RA3 Radiation levels that IMPEDE access Added defined term "IMPEDE" to clarify meaning.access to equipment necessary to equipment necessary for normal for normal plant operations, plant operations, cooldown or cooldown or shutdown shutdown.MODE: All MODE: All NEI EX. NEI Example EAL Wording BNP BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 Dose rate greater than 15 mR/hr RA3.1 Dose rate > 15 mR/hr in either of the No other site-specific areas requiring continuous occupancy in ANY of the following areas: following areas: exist at BNP." Control Room

Control Room ARM Channel 1-1 is the installed Control Room ARM." Central Alarm Station (ARM Channel 1-1) The CAS does not have installed area radiation monitoring" (other site-specific

° Central Alarm Station (by survey) and thus must be determined by survey.areas/rooms) 2 An UNPLANNED event results RA3.2 An UNPLANNED event results in The site-specific list of plant rooms or areas with entry-related in radiation levels that prohibit or radiation levels that prohibit or mode applicability are listed in Table R-2 for clarification. impede access to any of the IMPEDE access to any Table R-2 Added defined term "IMPEDE' to clarify meaning.following plant rooms or areas: rooms or areas (Note 5)(site-specific list of plant rooms or areas with entry-related mode applicability identified) Note If the equipment in the listed N/A Note 5 If the equipment in the listed None room or area was already area was already inoperable inoperable or out-of-service or out-of-service before the before the event occurred, then event occurred, then no 21 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP no emergency classification is emergency classification is warranted. warranted. Table R-2 Safe Operation & Shutdown Areas Room/Area Mode Applicability Reactor Building -17' North RHR Unit-1 & 2 3, 4, 5 Reactor Building -17' South RHR Unit-1 & 2 3,4,5 Reactor Building 20' East & West MCC Areas Unit-1 & 2 3, 4, 5 Reactor Building 20' Pipe Tunnel Unit-1 & 2 3,4,5 22 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BN P NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification Release of gaseous radioactivity RS1 Release of gaseous radioactivity None resulting in offsite dose greater resulting in offsite dose greater than than 100 mrem TEDE or 500 100 mrem TEDE or 500 mrem thyroid mrem thyroid CDE CDE MODE: All MODE: All NEI Ex. NEI Example EAL Wording BNP BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 Reading on ANY of the following RS1.1 In the absence of real-time dose The BNP radiation monitors that detect radioactivity effluent radiation monitors greater than assessment, reading on any Table R-1 release to the environment are listed in Table R-1. UE, Alert, the reading shown for 15 effluent radiation monitor > column SAE and GE thresholds for all BNP continuously monitored minutes or longer: "SAE" for > 15 min. gaseous release pathways are listed in Table R-1 to (site-specific monitor list and (Notes 1, 2, 3, 4) consolidate the information in a single location and, thereby, threshold values) simplify identification of the thresholds by the EAL-user.Added "In the absence of real-time dose assessment" to re-enforce Note 4.2 Dose assessment using actual RS1.2 Dose assessment using actual The site boundary is the site-specific receptor point.meteorology indicates doses meteorology indicates doses > 100 greater than 100 mrem TEDE or mrem TEDE or 500 mrem thyroid CDE 500 mrem thyroid CDE at or at or beyond the SITE BOUNDARY beyond (site-specific dose (Notes 3, 4)receptor point)3 Field survey results indicate RS1.3 Field survey results indicate EITHER The site boundary is the site-specific receptor point.EITHER of the following at or of the following at or beyond the SITE beyond (site-specific dose BOUNDARY: receptor point): e Closed window dose rates >" Closed window dose rates 100 mR/hr expected to continue greater than 100 mR/hr for > 60 min.expected to continue for 60 minutes or longer. 9 Analyses of field survey* Analyses of field survey samples indicate thyroid CDE >23 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP samples indicate thyroid 500 mrem for 60 min. of CDE greater than 500 inhalation. mrem for one hour of (Notes 1, 2)inhalation. Notes

The Emergency Director Note 1: The SEC should declare the The classification timeliness note has been standardized should declare the Site Area event promptly upon across the BNP EAL scheme by referencing the "time limit" Emergency promptly upon determining that time limit specified within the EAL wording.determining that the has been exceeded, or will applicable time has been likely be exceeded.exceeded, or will likely be Note 2: If an ongoing release is exceeded.

detected and the release The classification timeliness note has been standardized

If an ongoing release is start time is unknown, across the BNP EAL scheme by referencing the "time limit" detected and the release start assume that the release specified within the EAL wording.time is unknown, assume that duration has exceeded the the release duration has specified time limit.exceeded 15 minutes. Note 3: If the effluent flow past an* If the effluent flow past an effluent monitor is known to None effluent monitor is known to have stopped, indicating that have stopped due to actions the release path is isolated, to isolate the release path, the effluent monitor reading then the effluent monitor is no longer VALID for reading is no longer valid for classification purposes.classification purposes.

Note 4: The pre-calculated effluent* The pre-calculated effluent monitor values presented in monitor values presented in EALs RA1.1, RS1.1 and Incorporated site-specific EAL numbers associated with EAL #1 should be used for RG1.1 should be used for generic EAL#1.emergency classification emergency classification assessments until the results assessments until the results from a dose assessment from a dose assessment using actual meteorology are using actual meteorology are available. available. BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification AS2 Spent fuel pool level at (site- RS2 Spent fuel pool level at the top of the Top of the fuel racks is the site specific Level 3.specific Level 3 description) fuel racks 24 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP MODE: All NEI Ex. NEI Example EAL Wording EALBNP BNP EAL Wording Difference/Deviation Justification 1 Lowering of spent fuel pool level RS2.1 Lowering of spent fuel pool level to Post-Fukushima order EA-12-051 required the installation of to (site-specific Level 3 value) < 95 ft. 3 in. ele. reliable SFP level indication capable of identifying normal level (Level 1 -116 ft. 1 in. ele.), SFP level 10 ft. above the top of the fuel racks (Level 2 -105 ft. 3 in. ele.) and SFP level at the top of the fuel racks (Level 3 -95 ft. 3 in. ele.).25 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification AG1 Release of gaseous radioactivity RG1 Release of gaseous radioactivity None resulting in offsite dose greater resulting in offsite dose greater than 1,000 mrem TEDE or than 1,000 mrem TEDE or 5,000 5,000 mrem thyroid CDE. mrem thyroid CDE MODE: All MODE: All NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 Reading on ANY of the following RG1.1 In the absence of real-time dose The BNP radiation monitors that detect radioactivity effluent release radiation monitors greater than assessment, reading on any Table to the environment are listed in Table R-1. UE, Alert, SAE and GE the reading shown for 15 R-1 effluent radiation monitor> thresholds for all BNP continuously monitored gaseous release minutes or longer: column "GE" for > 15 min. (Notes pathways are listed in Table R-1 to consolidate the information in a (site-specific monitor list and 1,2, 3, 4) single location and, thereby, simplify identification of the thresholds threshold values) by the EAL-user.Added "In the absence of real-time dose assessment" to re-enforce Note 4.2 Dose assessment using actual RG1.2 Dose assessment using actual The site boundary is the site-specific receptor point.meteorology indicates doses meteorology indicates doses >greater than 1,000 mrem TEDE 1000 mrem TEDE or or 5,000 mrem thyroid CDE at 5000 mrem thyroid CDE at or or beyond (site-specific dose beyond the SITE BOUNDARY receptor point). (Notes 3, 4)3 Field survey results indicate RG1.3 Field survey results indicate The site boundary is the site-specific receptor point.EITHER of the following at or EITHER of the following at or beyond (site-specific dose beyond the SITE BOUNDARY: receptor point):

Closed window dose rates >" Closed window dose rates 1000 mR/hr expected to greater than 1,000 mR/hr continue for > 60 min.expected to continue for 60 minutes or longer. e Analyses of field survey* Analyses of field survey samples indicate thyroid CDE samples indicate thyroid CDE > 5000 mrem for 60 min. of 26 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP greater than 5,000 mrem for inhalation.

one hour of inhalation. (Notes 1, 2)Notes I* The Emergency Director should declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes." The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 5 The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. The classification timeliness note has been standardized across the BNP EAL scheme by referencing the "time limit" specified within the EAL wording.The classification timeliness note has been standardized across the BNP EAL scheme by referencing the "time limit" specified within the EAL wording.None Incorporated site-specific EAL numbers associated with generic EAL#1.27 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification AG2 Spent fuel pool level cannot be RG2 Spent fuel pool level cannot be Top of the fuel racks is the site specific Level 3.restored to at least (site-specific restored to at least the top of the fuel Level 3 description) for 60 racks for 60 minutes or longer minutes or longer MODE: All MODE: All NEI Ex. NEI Example EAL Wording BNP BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 Spent fuel pool level cannot be RG2.1 Spent fuel pool level cannot be Post-Fukushima order EA-12-051 required the installation of restored to at least (site-specific restored > 95 ft. 3 in. ele. for > 60 min. reliable SFP level indication capable of identifying normal level Level 3 value) for 60 minutes or (Note 1) (Level 1 -116 ft. 1 in. ele.), SFP level 10 ft. above the top of longer the fuel racks (Level 2 -105 ft. 3 in. ele.) and SFP level at the top of the fuel racks (Level 3 -95 ft. 3 in. ele.).">" is the same as "to at least" Note The Emergency Director should Note 1: The SEC should declare the The classification timeliness note has been standardized declare the General Emergency event promptly upon across the BNP EAL scheme by referencing the "time limit" promptly upon determining that determining that time limit specified within the EAL wording 60 minutes has been exceeded, has been exceeded, or will or will likely be exceeded. likely be exceeded.28 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Category C Cold Shutdown / Refueling System Malfunction 29 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording lC#(s) BNP IC Wording Difference/Deviation Justification CUl UNPLANNED loss of (reactor CUl UNPLANNED loss of RPV None vessel/RCS [PWR] or RPV inventory for 15 minutes or[BWR]) inventory for 15 minutes longer or longer. MODE: 4 -Cold Shutdown, 5 -MODE: Cold Shutdown, Refuel Refueling NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 UNPLANNED loss of reactor CU1.1 UNPLANNED loss of reactor None coolant results in (reactor coolant results in RPV water vessel/RCS [PWR] or RPV level less than a required lower[BWR]) level less than a limit for > 15 min. (Note 1)required lower limit for 15 minutes or longer.2 a. (Reactor vessel/RCS [PWR] CU1.2 RPV level cannot be monitored Site-specific applicable sumps and tanks are listed in Table C-1 to or RPV [BWRJ) level cannot AND improve the readability of the EAL.be monitored. UNPLANNED increase in any The phrase "due to a loss of RPV inventory" has been added to the Table C-i sump or tank levels BNP EAL for clarification. This wording implements the intent of the NEI EAL basis which states" "Sump and/or tank level changes must b. UNPLANNED increase in be evaluated against other potential sources of water flow to ensure (site-specific sump and/or they are indicative of leakage from the RCS." tank) levels. Although "Visual Observation" in Table C-1 is neither a sump nor tank, it is included in order to implement the intent of the NEI basis which states: "...operators may determine that an inventory loss is occurring by observing changes..." Note The Emergency Director should N/A Note 1: The SEC should The classification timeliness note has been standardized across the declare the Unusual Event declare the event BNP EAL scheme by referencing the "time limit" specified within the promptly upon determining that promptly upon EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded. limit has been exceeded, or will likely I 30 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP be exceeded.Table C-1 Sumps & Tanks" Drywell Floor Drain Sump* Drywell Equipment Drain Sump" RB Floor Drain Sumps" RB Equipment Drain Sumps* Torus" Visual Observation 31 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification CU2 Loss of all but one AC power CU2 Loss of all but one AC power None source to emergency buses for source to emergency buses for 15 minutes or longer. 15 minutes or longer.MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -Refueling, Defueled Refuel, D -Defueled NEI Ex. NEI Example EAL Wording BNP BNP EAL Wording Difference/Deviation Justification EAL # EAL #a. AC power capability to (site- CU2.1 AC power capability to 4 KV essential buses El (E3) and E2(E4) are the site-specific specific emergency buses) is Emergency 4 KV Buses El (E3) emergency buses.reduced to a single power and E2(E4) reduced to a single Because 2 RHR pumps on each unit are powered from the source for 15 minutes or power source for > 15 min. unaffected unit, the words "unit-specific" have been added to clarify longer. (Note 1) that the cross-connected RHR pump power cannot be credited as AND AND an AC power source relative to this EAL b. Any additional single power Any additional single power source failure will result in source failure will result in loss of loss of all AC power to all unit-specific AC power to SAFETY SYSTEMS. SAFETY SYSTEMS Note The Emergency Director should N/A Note 1: The SEC should The classification timeliness note has been standardized across the declare the Unusual Event declare the event BNP EAL scheme by referencing the "time limit" specified within the promptly upon determining that promptly upon EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded. limit has been exceeded, or will likely be exceeded.32 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification CU3 UNPLANNED increase in RCS CU3 UNPLANNED increase in RCS None temperature temperature MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -Refueling Refuel NEI Ex. BNP EA E NEI Example EAL Wording EAL BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 UNPLANNED increase in RCS CU3.1 UNPLANNED increase in RCS 212°F is the site-specific Tech. Spec. cold shutdown temperature temperature to greater than (site- temperature to > 212°F due to limit.specific Technical Specification loss of decay heat removal Added the phrase "due to loss of decay heat removal capability" to cold shutdown temperature limit) capability clarify that the increase in temperature is related to such capability as specified in the generic bases.2 Loss of ALL RCS temperature CU3.2 Loss of all RCS temperature and None and (reactor vessel/RCS [PWR] RPV level indication for > 15 min.or RPV [BWR]) level indication (Note 1)for 15 minutes or longer.Note The Emergency Director should N/A Note 1: The SEC should declare The classification timeliness note has been standardized across the declare the Unusual Event the event promptly upon BNP EAL scheme by referencing the "time limit" specified within the promptly upon determining that determining that time EAL wording.15 minutes has been exceeded, limit has been exceeded, or will likely be exceeded or will likely be exceeded.33 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification CU4 Loss of Vital DC power for 15 CU4 Loss of Vital DC power for 15 None minutes or longer. minutes or longer.MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -Refueling Refuel NEI Ex. NEI Example EAL Wording BNP BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 Indicated voltage is less than CU4.1 < 105 VDC bus voltage 105 VDC is the site-specific minimum vital DC bus design voltage (site-specific bus voltage value) indications on Technical per Technical Specification 3.8.4.on required Vital DC buses for Specification required 125 VDC 15 minutes or longer, buses for > 15 min. (Note 1)Note The Emergency Director should N/A Note 1: The SEC should declare The classification timeliness note has been standardized across declare the Unusual Event the event promptly upon the BNP EAL scheme by referencing the "time limit" specified promptly upon determining that determining that time limit has within the EAL wording.15 minutes has been exceeded, been exceeded, or will likely be or will likely be exceeded. exceeded.34 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BN P NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification CU5 Loss of all onsite or offsite CU5 Loss of all onsite or offsite None communications capabilities, communications capabilities. MODE: Cold Shutdown, MODE: 4- Cold Shutdown, 5 -Refueling, Defueled Refuel, D -Defueled NEI Ex. NEI Example EAL Wording BNP BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 Loss of ALL of the following CU5.1 Loss of all Table C-4 onsite Example EALs #1, 2 and 3 have been combined into a single EAL onsite communication methods: communication methods for simplification of presentation.(site specific list of OR Table C-4 provides a site-specific list of onsite, offsite (ORO) and communications methods) NRC communications methods.Loss of all Table C-4 offsite 2 Loss of ALL of the following ORO communication methods communications methods: (site specific list of OR communications methods) Loss of all Table C-4 NRC 3 Loss of ALL of the following NRC communication methods communications methods: (site specific list of communications methods)35 of 107 EAL Comparison Matrix OSSE Project #13-0703 BNP Table C-4 Communication Methods System Onsite Offsite NRC Public Address System X PBX Telephone System X X X Corporate Telephone System X X X Commercial Telephones X X X Satellite Phones X X Cellular Phones X X NRC emergency Telecommunications X System 36 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification CA1 Loss of (reactor vessel/RCS CA1 Loss of RPV inventory None[PWR] or RPV [BWR]) inventory MODE: 4 -Cold Shutdown, 5 -MODE: Cold Shutdown, Refuel Refueling NEI Ex. BNP EA E NEI Example EAL Wording EAL BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 Loss of (reactor vessel/RCS CA1.1 Loss of RPV inventory as 105 in. is the site-specific level corresponding to the Level 2 trip[PWR] or RPV [BWR]) inventory indicated by RPV water level < setpoint.as indicated by level less than 105 in. above TAF (Level 2)(site-specific level).2 a. (Reactor vessel/RCS [PWRJ CA1.2 RPV level cannot be monitored Site-specific applicable sumps and tanks are listed in Table C-1 to or RPV [BWR]) level cannot for a 15 min. (Note 1) improve the readability of the EAL.be monitored for 15 minutes AND The phrase "due to a loss of RPV inventory" has been added to the or longer UBNP EAL for clarification. This wording implements the intent of the AND Table C-1 sump or tank levels NEI EAL basis which states" "Sump and/or tank level changes must b. UNPLANNED increase in due to a loss of RPV inventory be evaluated against other potential sources of water flow to ensure (site-specific sump and/or they are indicative of leakage from the RCS." tank) levels due to a loss of Although "Visual Observation" in Table C-1 is neither a sump nor (reactor vessel/RCS [PWR] tank, it is included in order to implement the intent of the NEI basis or RPV [BWR]) inventory. which states: "...operators may determine that an inventory loss is occurring by observing changes..." Note The Emergency Director should N/A Note 1: The SEC should declare The classification timeliness note has been standardized across the declare the Alert promptly upon the event promptly upon BNP EAL scheme by referencing the "time limit" specified within the determining that 15 minutes has determining that time limit has EAL wording.been exceeded, or will likely be been exceeded, or will likely be exceeded exceeded.37 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification CA2 Loss of all offsite and all onsite CA2 Loss of all offsite and all onsite None AC power to emergency buses AC power to emergency buses for 15 minutes or longer for 15 minutes or longer.MODE: Cold Shutdown, MODE: Cold Shutdown, Refuel, Refueling, Defueled Defueled NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 Loss of ALL offsite and ALL CA2.1 Loss of all offsite and all onsite 4 KV emergency buses El (E3) and E2(E4) are the site-specific onsite AC Power to (site- AC power capability to emergency buses.specific emergency buses) for Emergency 4 KV Buses El (E3)15 minutes or longer, and E2(E4) for > 15 min.(Note 1)Note The Emergency Director should N/A Note 1: The SEC should The classification timeliness note has been standardized across the declare the Unusual Event declare the event BNP EAL scheme by referencing the "time limit" specified within the promptly upon determining that promptly upon EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded. limit has been exceeded, or will likely be exceeded.38 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification CA3 Inability to maintain the plant in CA3 Inability to maintain the plant in None cold shutdown. cold shutdown.MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -Refueling Refuel NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification UNPLANNED increase in RCS CA3.1 UNPLANNED increase in RCS EALs #1 and #2 combined into a single EAL based on user temperature to greater than temperature to > 212'F for > preference.(site-specific Technical Table C-3 duration (Note 1) 212OF is the site-specific Tech. Spec. cold shutdown temperature Specification cold shutdown OR limit.temperature limit) for greater than the duration specified in UNPLANNED RPV pressure Table C-3 is the site-specific implementation of the generic RCS the following table. increase > 10 psig due to a Reheat Duration Threshold table.loss of RCS cooling 10 psig is the site-specific pressure increase readable by Control 2 UNPLANNED RCS pressure Room indications. increase greater than (site-specific pressure reading). Added the phrase "due to loss of RCS cooling" to clarify that the (This EAL does not apply during increase in temperature is related to such loss as specified in the water-solid plant conditions. generic bases.[PWFR])Note The Emergency Director should N/A Note 1: The SEC should The classification timeliness note has been standardized across the declare the Unusual Event declare the event BNP EAL scheme by referencing the "time limit" specified within the promptly upon determining that promptly upon EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded. limit has been exceeded, or will likely be exceeded.39 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Table: RCS Heat-up Duration Thresholds RCS Status Containment Closure Status Heat-up Duration Intact (but not at reduced Not applicable 60 minutes*inventory [PWR])Not intact (or at reduced Established 20 minutes*inventory [PWR]) Not Established 0 minutes* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable. Table C-3: RCS Heat-up Duration Thresholds RCS Status Containment Closure Status Heat-up Duration Intact N/A 60 min.*established 20 min.*Not intact not established 0 min.* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable. 40 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification CA6 Hazardous event affecting a CA6 Hazardous event affecting a None SAFETY SYSTEM needed for SAFETY SYSTEM needed for the current operating mode. the current operating mode.MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -Refueling Refuel NEI Ex. NEI Example EAL Wording BNP BNP EAL Wording Difference/Deviation Justification EAL # EAL #a. The occurrence of ANY of the CA6.1 The occurrence of any Table C- The hazardous events have been listed in Table C-5 to improve the following hazardous events: 5 hazardous event readability of the BNP EAL." Seismic event AND The NEI list of hazardous events includes all BNP hazardous events.(earthquake) EITHER of the following: No additional hazardous events could be identified." Internal or external flooding event

Event damage has" High winds or tornado caused indications of strike degraded performance in" FIRE at least one train of a" EXPLOSION SAFETY SYSTEM" (site-specific hazards) needed for the current" Other events with similar operating mode hazard characteristics as detemind b th Shft* The event has caused determined by the Shift VISIBLE DAMAGE to a Manager SAFETY SYSTEM b. EITHER of the following:

component or structure 1. Event damage has caused operating mode indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.OR 41 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.Table C-5 Hazardous Events* Seismic event (earthquake)

Internal or external FLOODING event* High winds or tornado strike* FIRE* EXPLOSION o Other events with similar hazard characteristics as determined by the Shift Manager 42 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification CS1 Loss of (reactor vessel/RCS CS1 Loss of RPV inventory affecting None[PWR] or RPV [BWR]) inventory core decay heat removal affecting core decay heat capability removal capability.

MODE: 4 -Cold Shutdown, 5 -MODE: Cold Shutdown, Refuel Refueling NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 a. CONTAINMENT CLOSURE CS1.1 CONTAINMENT CLOSURE not 45 in. is the RPV water level that corresponds to the Level 3 trip not established, established setpoint.AND AND b. (Reactor vessel/RCS [PWR] RPV level < 45 in. (Level 3) Note that for consistency with plant terminology, BNPs Level 3 is or RPV [BWRF) level less than equivalent to the NEI 99-01 Level 1.(site-specific level).2 a. CONTAINMENT CLOSURE CS1.2 CONTAINMENT CLOSURE None established, established AND AND b. (Reactor vessel/RCS [PWR] RPV water level < TAF or RPV [BWR]) level less than (site-specific level).3 a. (Reactor vessel/RCS [PWF] CS1.3 RPV water level cannot be Site-specific applicable sumps and tanks are listed in Table C-1 to or RPV [BWR]) level cannot monitored for > 30 min. (Note 1) improve readability of the EAL.be monitored for 30 minutes AND The phrase "due to a loss of RPV inventory" has been added to the or longer. CBNP EAL for clarification. This wording implements the intent of the b. D Core uncovery is indicated by ANY of the following:

UNPLANNED increase in they are indicative of leakage from the RCS." any Table C-1 sump or II 43 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP" (Site-specific radiation monitor) reading greater than (site-specific value)" Erratic source range monitor indication

[PWR]" UNPLANNED increase in (site-specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery" (Other site-specific indications) tank levels due to a loss of RPV inventory UNPLANNED increase in ARM Channel 28 Between Reactor and Fuel Pool > 1000 mR/hr Although "Visual Observation" in Table C-1 is neither a sump nor tank, it is included in order to implement the intent of the NEI basis which states: "...operators may determine that an inventory loss is occurring by observing changes..." ARM Channel 28 Between Reactor and Fuel Pool is located on the Refuel Floor in the Reactor Building and is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations. If this radiation monitor reaches and exceeds the alarm setpoint of 1,000 mr/hr, a loss of inventory with potential to uncover the core is likely to have occurred.Note The Emergency Director should N/A Note 1: The SEC should declare The classification timeliness note has been standardized across the declare the Site Area the event promptly upon BNP EAL scheme by referencing the "time limit" specified within the Emergency promptly upon determining that time EAL wording.determining that 30 minutes has limit has been been exceeded, or will likely be exceeded, or will likely exceeded be exceeded.44 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification CG1 Loss of (reactor vessel/RCS CG1 Loss of RPV inventory affecting None[PWR] or RPV [BWR]) inventory fuel clad integrity with affecting fuel clad integrity with Containment challenged containment challenged MODE: 4 -Cold Shutdown, 5 -MODE: Cold Shutdown, Refuel Refueling NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 a. (Reactor vessel/RCS [PWR] or CG1.1 RPV water level < TAF for > 30 6% hydrogen concentration in the presence of oxygen represents RPV [BWR]) level less than mi. (Note 1) an explosive mixture in primary containment.(site-specific level) for 30 minutes or longer. AND The Max Safe Operating Radiation Levels are the highest value of these parameters at which neither: (1) equipment necessary for AND Any Containment Challenge the safe shutdown of the plant will fail, nor (2) personnel access indication, Table C-2 necessary for the safe shutdown of the plant will be precluded.

b. ANY indication from the These are the site-specific secondary containment radiation Containment Challenge Table monitor readings and are listed in EOP-03-SCCP Table 3.(see below).2 a. (Reactor vessel/RCS

[PWR] or CG1.2 RPV water level cannot be Site-specific applicable sumps and tanks are listed in Table C-1 to RPV [BWR]) level cannot be monitored for > 30 min. (Note 1) improve the readability of the EAL.monitored for 30 minutes or AND The phrase "due to a loss of RPV inventory" has been added to the ong. Core uncovery is indicated by BNP EAL for clarification. This wording implements the intent of EITHER of the following: the NEI EAL basis which states" "Sump and/or tank level changes b. Core uncovery is indicated by must be evaluated against other potential sources of water flow to ANY of the following: 0 UNPLANNED increase in ensure they are indicative of leakage from the RCS." (Site-specific radiation any Table C-1 sump or Although "Visual Observation" in Table C-1 is neither a sump nor monitor) readin ater tank levels due to a loss of tank, it is included in order to implement the intent of the NEI basis monitor) reading greater RPV inventory which states: "...operators may determine that an inventory loss is than (site-specific value)" Erratic source range 0 UNPLANNED increase in occurring by observing changes..."__rrtcouc range_____ _ARM Channel 28 Between ARM Channel 28 Between Reactor and Fuel Pool is located on the 45 of 107 EAL Comparison Matrix OSSE Project #13-0703 BNP Reactor and Fuel Pool>monitor indication [PWRI]* UNPLANNED increase in (site-specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery* (Other site-specific indications) AND c. ANY indication from the Containment Challenge Table (see below).Reactor and Fuel Pool >1000 mR/hr AND Any Containment Challenge indication, Table C-2 Refuel Floor in the Reactor Building and is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations. If this radiation monitor reaches and exceeds the alarm setpoint of 1,000 mr/hr, a loss of inventory with potential to uncover the core is likely to have occurred.6% hydrogen concentration in the presence of oxygen represents an explosive mixture in primary containment. The Max Safe Operating Radiation Levels are 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. These are the site-specific secondary containment radiation monitor readings and are listed in EOP Table 14.+ I. *4 4 Note The Emergency Director should declare the General Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.N/A N/A Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.The classification timeliness note has been standardized across the BNP EAL scheme by referencing the "time limit" specified within the EAL wording.Note 6 implements the asterisked note associated with the generic Containment Challenge table.46 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Containment Challenge Table" CONTAINMENT CLOSURE not established*

(Explosive mixture) exists inside containment" UNPLANNED increase in containment pressure" Secondary containment radiation monitor reading above (site-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.Table C-2 Containment Challenge Indications
CONTAINMENT CLOSURE not established (Note 6)* Primary Containment hydrogen concentration

> 6%* UNPLANNED rise in PC pressure* Exceeding one or more Secondary Containment Control Maximum Safe Operating Area Radiation Levels (EOP-03-SCCP Table 3)Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required 47 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Category D Permanently Defueled Station Malfunction 48 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification PD-AU1 Recognition Category D N/A N/A NEI Recognition Category PD ICs and EALs are applicable only to PD-AU2 Permanently Defueled Station permanently defueled stations. BNP is not a defueled station.PD-SU1 PD-HU1 PD-HU2 PD-HU3 PD-AA1 PD-AA2 PD-HA1 PD-HA3 49 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Category E Independent Spent Fuel Storage Installation (ISFSI)50 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification E-HU1 Damage to a loaded cask EU1 Damage to a loaded cask None CONFINEMENT BOUNDARY CONFINEMENT BOUNDARY MODE: All MODE: All NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 Damage to a loaded cask EU1.1 Damage to a loaded canister 1400 mrem/hr on the front surface, 10 mrem/hr on the door CONFINEMENT BOUNDARY as confinement boundary as centerline of the overpack and 20 mrem/hr on the end shield wall of indicated by an on-contact indicated by an on-contact the overpack represent 2 times the site-specific cask technical radiation reading greater than (2 radiation reading on the surface specification allowable levels per the ISFSI Technical Specifications times the site-specific cask of a loaded spent fuel cask > (CoC).specific technical specification any of the following: allowable radiation level) on the e 1,400 mrem/hr on the HSM-surface of the spent fuel cask. H front surface* 10 mrem/hr on the HSM-H door centerline

20 mrem/hr on the end shield wall exterior 51 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Category F Fission Product Barrier Degradation 52 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification FA1 Any Loss or any Potential Loss of FA1 Any loss or any potential loss of Deleted "Hot Standby" because BWRs do not have this operating either the Fuel Clad or RCS either Fuel Clad or RCS mode.barrier. MODE: 1 -Power Operation, 2 -MODE: Power Operation, Hot Startup, 3 -Hot Shutdown Standby, Startup, Hot Shutdown NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 Any Loss or any Potential Loss of FA1.1 Any loss or any potential loss of Table F-1 provides the fission product barrier loss and potential loss either the Fuel Clad or RCS either Fuel Clad or RCS barrier thresholds.

barrier. (Table F-1)53 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification FS1 Loss or Potential Loss of any two FS1 Loss or potential loss of any two Deleted "Hot Standby" because BWRs do not have this operating barriers barriers mode.MODE: Power Operation, Hot MODE: 1 -Power Operation, 2 -Standby, Startup, Hot Shutdown Startup, 3 -Hot Shutdown NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 Loss or Potential Loss of any two FS1.1 Loss or potential loss of any two Table F-1 provides the fission product barrier loss and potential loss barriers barriers thresholds. 54 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification FG1 Loss of any two barriers and FG1 Loss of any two barriers and loss Deleted "Hot Standby" because BWRs do not have this operating Loss or Potential Loss of third or potential loss of the third mode.barrier barrier MODE: Power Operation, Hot MODE: 1 -Power Operation, 2 -Standby, Startup, Hot Shutdown Startup, 3 -Hot Shutdown NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 Loss of any two barriers and FG1.1 Loss of any two barriers Table F-1 provides the fission product barrier loss and potential loss Loss or Potential Loss of third AND thresholds. barrier Loss or potential loss of the third barrier (Table F-i)55 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BWR Fuel Clad Fission Product Barrier Degradation Thresholds NEI BNP FPB NEI Threshold Wording FP BNP FPB Wording Difference/Deviation Justification FPB# FPB #(s)FC Loss RCS Activity FC Loss Primary coolant activity > 300 300 pCi/gm DEl-1 31 is the site-specific indication for this reactor 1 A. (Site-specific indications that D.2 pCi/gm DEl-131 dose coolant activity.reactor coolant activity is greater equivalent than 300 pCi/gm dose equivalent 1-131).FC Loss RPV Water Level FC Loss Primary Containment Flooding None 2 A. Primary containment flooding A.1 required required.FC Loss Not Applicable N/A N/A N/A 3 Not Applicable FC Loss Primary Containment FC Loss Drywell radiation > 2,000 R/hr 2,000 R/hr is the site-specific primary containment radiation monitor 4 Radiation D.1 reading associated with 2% fuel clad.A. Primary containment radiation monitor reading greater than (site-specific value).FC Loss Other Indications N/A N/A No other site-specific Fuel Clad Loss indication has been identified 5 for BNP.A. (site-specific as applicable) FC Loss Emergency Director FC Loss Any condition in the opinion of None 6 Judgment F.1 the Site Emergency Coordinator that indicates loss A. ANY condition in the opinion of the Fuel Clad barrier of the Emergency Director that indicates Loss of the Fuel Clad Barrier.56 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP NEI BNP NPI NEI Threshold Wording FP BNP FPB Wording Difference/Deviation Justification FPB# FPB #(s)FC RCS Activity N/A N/A N/A P-Loss Not Applicable 1 FC RPV Water Level FC RPV level cannot be restored None P-Loss A. RPV water level cannot be P-Loss and maintained > TAF or cannot 2 restored and maintained above A.1 be determined (site-specific RPV water level corresponding to the top of active fuel) or cannot be determined. FC Not Applicable N/A N/A N/A P-Loss Not Applicable 3 FC Primary Containment N/A N/A N/A P-Loss Radiation 4 Not Applicable FC Other Indications N/A N/A No other site-specific Fuel Clad Potential Loss indication has been P-Loss identified for BNP.A. (site-specific as applicable) 5 FC Emergency Director FC Any condition in the opinion of None P-Loss Judgment P-Loss the Site Emergency Coordinator 6 A. Any condition in the opinion F.1 that indicates potential loss of of the Emergency Director that the fuel clad barrier indicates Potential Loss of the Fuel Clad Barrier.57 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BWR RCS Fission Product Barrier Degradation Thresholds NEI BNP FPB FPB NEI IC Wording #(s BNP FPB Wording Difference/Deviation Justification FPB# #(s)RCS Primary Containment RCS Loss Primary Containment 1.7 psig is the site-specific primary containment pressure Loss Pressure C.1 pressure > 1.7 psig due to corresponding to the drywell high pressure scram and isolation 1 A. Primary containment pressure RCS leakage setpoint.greater than (site-specific value)due to RCS leakage.RCS RPV Water Level RCS Loss RPV level cannot be restored None Loss and maintained > TAF or A. RPV water level cannot be A.1 anntb ined 2 restored and maintained above (site-specific RPV water level corresponding to the top of active fuel) or cannot be determined. RCS RCS Leak Rate RCS Loss UNISOLABLE break outside Main Steam Line, HPCI Steam Line, RCIC Steam Line, RWCU, Loss A. UNISOLABLE break in ANY B.1 Primary Containment in any of and Feedwater are the site-specific systems with potential for high 3 of the following: (site-specific the following: energy line breaks.systems with potential for high-

Main Steam Line Added "outside Primary Containment" to clarify intent.energy line breaks)
HPCI Steam Line OR 0 RCIC Steam Line B. Emergency RPV 0 RWCU Depressurization.
Feedwater RCS Loss Emergency RPV None B.2 Depressurization is required 58 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP NEI BNP FPB FPB NEI IC Wording #(s BNP FPB Wording Difference/Deviation Justification FPB# 9#(s)RCS Primary Containment RCS Loss Drywell radiation

> 27 R/hr 27 R/hr is the site-specific primary containment radiation monitor Loss Radiation DA with reactor shutdown reading associated with normal coolant noble gas and iodine 4 A. Primary containment inventory released instantaneously into containment. radiation monitor reading greater than (site-specific value).RCS Other Indications N/A N/A No other site-specific RCS Loss indication has been identified for Loss BNP.A. (site-specific as applicable) 5 RCS Emergency Director Judgment RCS Loss Any condition in the opinion None Loss A. ANY condition in the opinion F.1 of the Site Emergency Coordinator that indicates loss 6 of the Emergency Director that o h C are indicates Loss of the RCS Barrier.RCS Primary Containment N/A N/A N/A P-Loss 1 Pressure Not Applicable RCS RPV Water Level N/A N/A N/A P-Loss 2 Not Applicable 59 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP NEI BNP FPB FPB NEI IC Wording #(s BNP FPB Wording Difference/Deviation Justification FPB# #(s)RCS RCS Leak Rate RCS UNISOLABLE primary system Reference to EOP Tables 1 and 3 has been added for clarification. P-Loss 3 A. UNISOLABLE primary P-Loss leakage that results in system leakage that results in B.1 exceeding EITHER of the exceeding EITHER of the following: following: One or more Secondary 1. Max Normal Operating Containment area 1.mpraxNrmer radiation Maximum Normal Temperature Operating Limits (OEOP-OR 03-SCCP Table 3)2. Max Normal Operating Area One or more Secondary Radiation Level. Containment area temperature Maximum Normal Operating Limits (OEOP-03-SCCP Table 1)RCS Primary Containment N/A N/A N/A P-Loss 4 Radiation Not Applicable RCS Other Indications N/A N/A No other site-specific RCS Potential Loss indication has been P-Loss 5 A. (site-specific as applicable) identified for BNP.RCS Emergency Director Judgment RCS Any condition in the opinion of None P-Loss 6 A. ANY condition in the opinion P-Loss the Site Emergency Coordinator that indicates of the Emergency Director that F.1 potential loss of the RCS barrier indicates Potential Loss of the RCS Barrier.60 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BWR Containment Fission Product Barrier Degradation Thresholds NEI NEI IC Wording BNP BNP FPB Wording Difference/Deviation Justification FPB# FPB #(s)PC Loss Primary Containment Conditions PC Loss UNPLANNED rapid drop in Primary None Containment pressure following Primary 1 A. UNPLANNED rapid drop in primary C.1 Containment pressure rise containment pressure following primary containment pressure rise OR PC Loss Primary Containment pressure None B. Primary containment pressure C.2 response not consistent with LOCA response not consistent with LOCA conditions conditions. PC Loss RPV Water Level N/A N/A N/A 2 Not Applicable PC Loss Primary Containment Isolation PC Loss UNISOLABLE direct downstream None 3 Failure E.1 pathway to the environment exists after A. UNISOLABLE direct downstream Primary Containment isolation signal pathway to the environment exists after primary containment isolation signal OR PC Loss Intentional Primary Containment venting None B. Intentional primary containment E.2 per EOPs venting per EOPs OR 61 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP NEI BNP FPB NEI IC Wording FP BNP FPB Wording Difference/Deviation Justification FPB# FPB #(s)C. UNISOLABLE primary system PC Loss UNISOLABLE primary system leakage Reference to EOP Tables 1 and 3 has been added for leakage that results in exceeding that results in exceeding EITHER of the clarification. EITHER of the following: B.1 following:

1. Max Safe Operating N One or more Secondary Temperature.

Containment area radiation OR Maximum Safe Operating Limits 2. Max Safe Operating Area (OEOP-03-SCCP Table 3)Radiation Level. 0 One or more Secondary Containment area temperature Maximum Safe Operating Limits (OEOP-03-SCCP Table 1)PC Loss Primary Containment Radiation N/A N/A N/A 4 Not Applicable PC Loss Other Indications N/A N/A No other site-specific Containment Loss indication has 5 A. (site-specific as applicable) been identified for BNP.PC Emergency Director Judgment PC Any condition in the opinion of the Site The NEI term "containment" has been changed to Loss ANY condition in the opinion of the Loss Emergency Coordinator that indicates "Primary Containment" for clarification. 6 Emergency Director that indicates Loss F.1 loss of the Primary Containment barrier of the Containment Barrier.PC Primary Containment Conditions PC Primary Containment Pressure 62 psig is the maximum BNP primary containment P-Loss A. Primary containment pressure P-Loss > 62 psig pressure allowed by design.1 greater than (site-specific value) B.1 OR PC Deflagration concentrations exist inside 6% H 2 and > 5% 02 in either the drywell or B. (site-specific explosive mixture) P-Loss PC (H 2 -> 6% AND 02 -- 5%) suppression chamber are the site-specific indications exists inside primary containment B.2 of an explosive mixture inside primary containment. 62 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP NEI NEI IC Wording BNPBNP FPB Wording Difference/Deviation Justification FPB# FPB #(s)OR PC Heat Capacity Temperature Limit None C. HCTL exceeded. P-Loss (HCTL) exceeded B.3 PC RPV Water Level PC Primary Containment Flooding required None P-Loss P-Loss A. Primary containment flooding 2 required. A.1 PC Primary Containment Isolation N/A N/A N/A P-Loss Failure 3 Not Applicable PC Primary Containment Radiation PC Drywell radiation > 20,000 R/hr 20,000 R/hr is the site-specific primary containment P-Loss A. Primary containment radiation P-Loss radiation monitor reading associated with 20% fuel 4 monitor reading greater than (site- D.1 clad failure released instantaneously into containment. specific value).PC Other Indications N/A N/A No other site-specific Containment Potential Loss P-Loss indication has been identified for BNP.A. (site-specific as applicable) 5 PC Emergency Director Judgment PC Any condition in the opinion of the Site The NEI term "containment" has been changed to P-Loss A. ANY condition in the opinion of the P-Loss Emergency Coordinator that indicates "Primary Containment" for clarification. A. AY cndiion n te oinio ofthepotential loss of the Primary 6 Emergency Director that indicates F.1 Containment barrier Potential Loss of the Containment Barrier.63 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Category H Hazards and Other Conditions Affecting Plant Safety 64 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification HU1 Confirmed SECURITY HU1 None ConDITION oECURIY ta Confirmed SECURITY CONDITION or threat.CONDITION or threat MODE: All MODE: All NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 A SECURITY CONDITION that HU1.1 A SECURITY CONDITION that does not involve Security Shift Supervision is the site-specific does not involve a HOSTILE a HOSTILE ACTION as reported by the Security security shift supervision. ACTION as reported by the (site- Shift Supervision specific security shift supervision). 2 Notification of a credible security HU1.2 Notification of a credible security threat directed None threat directed at the site. at the site 3 A validated notification from the HU1.3 A validated notification from the NRC providing None NRC providing information of an information of an aircraft threat aircraft threat.65 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification HU2 Seismic event greater than OBE HU2 Seismic event greater than OBE None levels levels MODE: All MODE: All NEI Ex. NEI Example EAL Wording BNP BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 Seismic event greater than HU2.1 Seismic event > OBE per OAOP- OAOP-1 3.0 provides guidance for determining OBE exceedance. Operating Basis Earthquake 13.0 (OBE) as indicated by: (site-specific indication that a seismic event met or exceeded OBE limits)66 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification HU3 Hazardous event. HU3 Hazardous event None MODE: All MODE: All NEI Ex. BNP EA NEI Example EAL Wording EAL BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 A tornado strike within the HU3.1 A tornado strike within the None PROTECTED AREA. PROTECTED AREA 2 Internal room or area flooding of a HU3.2 Internal room or area FLOODING None magnitude sufficient to require of a magnitude sufficient to manual or automatic electrical require manual or automatic isolation of a SAFETY SYSTEM electrical isolation of a SAFETY component needed for the current SYSTEM component needed for operating mode. the current operating mode 3 Movement of personnel within the HU3.3 Movement of personnel within the Replaced the term "offsite" with "external to the PROTECTED PROTECTED AREA is impeded PROTECTED AREA is AREA". The term "offsite has specific meaning not consistent with due to an offsite event involving IMPEDED due to an event the intent of the EAL.hazardous materials (e.g., an external to the PROTECTED offsite chemical spill or toxic gas AREA involving hazardous release). materials (e.g., an offsite chemical spill or toxic gas release)4 A hazardous event that results in HU3.4 A hazardous event that results in Added reference to Note 7.on-site conditions sufficient to onsite conditions sufficient to prohibit the plant staff from prohibit the plant staff from accessing the site via personal accessing the site via personal vehicles. vehicles (Note 7)5 (Site-specific list of natural or HU3.5 Intake Canal water level > +19 ft The high Intake Canal level is the highest remotely measurable technological hazard events) Mean Sea Level intake canal level. Otherwise the specified level would have been 67 of 107 EAL Comparison Matrix OSSE Project #13-0703 BNP OR based the plant design that Class I structures and engineered Intake Canal water level < -7.75 ft safety features systems are protected against still water flooding Mean Sea Level (elevation 22.0 feet).The minimum water level predicted for the Maximum Probable Hurricane is -7.5 feet Mean Sea Level under special case circumstances Note EAL #4 does not apply to routine N/A Note 7: This EAL does not traffic impediments such as fog, apply to routine traffic snow, ice, or vehicle breakdowns impediments such as or accidents. fog, snow, ice, or vehicle breakdowns or accidents. 68 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification HU4 FIRE potentially degrading the HU4 FIRE potentially degrading the None level of safety of the plant. level of safety of the plant MODE: All MODE: All NEI Ex. BNP EA NEI Example EAL Wording EAL BNP EAL Wording Difference/Deviation Justification EAL # EAL #a. A FIRE is NOT extinguished HU4.1 A FIRE is not extinguished within Site-specific plant rooms and areas are listed in Table H-1 to within 15-minutes of ANY of the 15 min. of any of the following improve the readability of the EAL.following FIRE detection FIRE detection indications (Note indications: 1): " Report from the field (i.e., e Report from the field (i.e., visual observation) visual observation)

Receipt of multiple (more
Receipt of multiple (more than 1) fire alarms or than 1) fire alarms or indications indications
Field verification of a single
Field verification of a single fire alarm fire alarm AND AND b. The FIRE is located within The FIRE is located within any ANY of the following plant rooms Table H-1 area or areas: (site-specific list of plant rooms or areas)2 a. Receipt of a single fire alarm HU4.2 Receipt of a single fire alarm Site-specific plant rooms and areas are listed in Table H-1 to (i.e., no other indications of a (i.e., no other indications of a improve the readability of the EAL.FIRE). FIRE)AND AND b. The FIRE is located within The fire alarm is indicating a 69 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP ANY of the following plant rooms FIRE within any Table H-1 area or areas: AND (site-specific list of plant rooms or The existence of a FIRE is not areas) verified within 30 min. of alarm AND receipt (Note 1)c. The existence of a FIRE is not verified within 30-minutes of alarm receipt.3 A FIRE within the plant or ISFSI HU4.3 A FIRE within the plant BNP does not have an ISFSI located outside the plant Protected[for plants with an ISFSI outside PROTECTED AREA not Area.the plant Protected Area] extinguished within 60 min. of the PROTECTED AREA not initial report, alarm or indication extinguished within 60-minutes of (Note 1)the initial report, alarm or indication.

4 A FIRE within the plant orlSFSI HU4.4 A FIRE within the plant None[for plants with an ISFSl outside PROTECTED AREA that the plant Protected Area] requires firefighting support by PROTECTED AREA that requires an offsite fire response agency to firefighting support by an offsite extinguish fire response agency to extinguish. Note Note: The Emergency Director N/A Note 1: The SEC should declare The classification timeliness note has been standardized across the should declare the Unusual Event the event promptly upon BNP EAL scheme by referencing the "time limit" specified within the promptly upon determining that determining that time EAL wording.the applicable time has been limit has been exceeded, exceeded, or will likely be or will likely be exceeded.exceeded. orwilllikelybeexceeded. 70 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Table H-1 Fire Areas* Reactor Building* Diesel Generator Building* Diesel 4-Day Tank Rooms* Service Water Building* Turbine Building* Control Building* CSTs* Diesel Fuel Oil Storage Tank 71 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification HU7 Other conditions exist which in the HU7 Other conditions existing that in None judgment of the Emergency the judgment of the Site Director warrant declaration of a Emergency Coordinator warrant (NO)UE declaration of a UE MODE: All MODE: All NEI Ex. NEBxmleELWrdnPA ENEI Example EAL Wording BNP BNP EAL Wording Difference/Deviation Justification EAL # EAL #Other conditions exist which in the HU7.1 Other conditions exist which in None judgment of the Emergency the judgment of the Site Director indicate that events are in Emergency Coordinator indicate progress or have occurred which that events are in progress or indicate a potential degradation of have occurred which indicate a the level of safety of the plant or potential degradation of the level indicate a security threat to facility of safety of the plant or indicate a protection has been initiated. No security threat to facility releases of radioactive material protection has been initiated. No requiring offsite response or releases of radioactive material monitoring are expected unless requiring offsite response or further degradation of safety monitoring are expected unless systems occurs. further degradation of safety systems occurs.72 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification HA1 HOSTILE ACTION within the HA1 HOSTILE ACTION within the None OWNER CONTROLLED AREA or OWNER CONTROLLED AREA airborne attack threat within 30 or airborne attack threat within 30 minutes. minutes MODE: All MODE: All NEI Ex. BNP EA E NEI Example EAL Wording EAL BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 A HOSTILE ACTION is occurring or HAl.1 A HOSTILE ACTION is The Security Shift Supervision is the site-specific security shift has occurred within the OWNER occurring or has occurred within supervision. CONTROLLED AREA as reported the OWNER CONTROLLED by the (site-specific security shift AREA as reported by the supervision). Security Shift Supervision 2 A validated notification from NRC of HA1.2 A validated notification from None NRC of an aircraft attack threat Nn an aircraft attack threat within 30 wti 30 mn of theeit mwithin 30 min. of the site 73 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification Gaseous release impeding HA5 Gaseous release IMPEDING Added defined term "IMPEDE" to clarify meaning.access to equipment necessary access to equipment necessary for normal plant operations, for normal plant operations, cooldown or shutdown. cooldown or shutdown MODE: All MODE: All NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification

a. Release of a toxic, HA5.1 Release of a toxic, corrosive, Plant rooms or areas with entry-related mode applicability are listed corrosive, asphyxiant or asphyxiant or flammable gas into in Table H-2 to improve the readability of the EAL.flammable gas into any of the any Table H-2 area Added defined term "IMPEDE" to clarify meaning.following plant rooms or areas: AND (site-specific list of plant rooms or areas with entry-related mode Entry into the area is prohibited or applicability identified)

IMPEDED (Note 5)AND b. Entry into the room or area is prohibited or impeded.Note Note: If the equipment in the N/A Note 5: If the equipment in the None listed room or area was already listed area was already inoperable or out-of-service inoperable or out-of-before the event occurred, then service before the event no emergency classification is occurred, then no warranted. emergency classification is warranted. 74 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Table H-2 Safe Operation & Shutdown Areas Room/Area Mode Applicability Reactor Building -17' North RHR Unit-1 & 2 3,4,5 Reactor Building -17' South RHR Unit-1 & 2 3,4,5 Reactor Building 20' East & West MCC Areas Unit-1 & 2 3,_4,_5 Reactor Building 20' Pipe Tunnel Unit-1 & 2 3,14,_5 75 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification HA6 Control Room evacuation HA6 Control Room evacuation None resulting in transfer of plant resulting in transfer of plant control to alternate locations, control to alternate locations MODE: All MODE: All NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 An event has resulted in plant HA6.1 An event has resulted in plant The Remote Shutdown Panels is the site-specific remote shutdown control being transferred from the control being transferred from the panels/local control stations.Control Room to (site-specific Control Room to the Remote remote shutdown panels and Shutdown Panels local control stations). 76 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification HA7 Other conditions exist which in the HA7 Other conditions exist that in the None judgment of the Emergency Director judgment of the Site Emergency warrant declaration of an Alert. Coordinator warrant declaration of an MODE: All Alert MODE: All NEI Ex. NEI Example EAL Wording BNP BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 Other conditions exist which, in the HA7.1 Other conditions exist which, in the None judgment of the Emergency Director, judgment of the Site Emergency indicate that events are in progress or Coordinator, indicate that events are in have occurred which involve an actual or progress or have occurred which involve potential substantial degradation of the an actual or potential substantial level of safety of the plant or a security degradation of the level of safety of the event that involves probable life plant or a security event that involves threatening risk to site personnel or probable life threatening risk to site damage to site equipment because of personnel or damage to site equipment HOSTILE ACTION. Any releases are because of HOSTILE ACTION. Any expected to be limited to small fractions releases are expected to be limited to of the EPA Protective Action Guideline small fractions of the EPA Protective exposure levels. Action Guideline exposure levels.77 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification HS1 HOSTILE ACTION within the HS1 HOSTILE ACTION within the None PROTECTED AREA PROTECTED AREA MODE: All MODE: All NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 A HOSTILE ACTION is occurring HS1.1 A HOSTILE ACTION is occurring or has The Security Shift Supervision is the site-specific security or has occurred within the occurred within the PROTECTED AREA shift supervision. PROTECTED AREA as reported as reported by the Security Shift by the (site-specific security shift Supervision supervision). 78 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification HS6 Inability to control a key safety HS6 Inability to control a key safety function None function from outside the Control from outside the Control Room Room. MODE: All MODE: All NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 a. An event has resulted in HS6.1 An event has resulted in plant control The Remote Shutdown Panels are the site-specific remote plant control being transferred being transferred from the Control Room shutdown panels/local control stations.from the Control Room to (site- to the Remote Shutdown Panel Deleted the word "control" after "reactivity" as it is specific remote shutdown panels AND redundant. and local control stations). AND Control of any of the following key 22.5 min. is the site specific time required to establish safety functions is not reestablished control (RCIC start) per Calculation No. BNP-E-9.007 ASSD b. Control of ANY of the within 22.5 min. (Note 1): Manual Action Feasibility. following key safety functions is not reestablished within (site-specific number of minutes). 9 RPV water level" Reactivity control

RCS heat removal" Core cooling [PWR] / RPV water level [BWR]* RCS heat removal N/A N/A N/A Note 1: The SEC should declare the Added the standard Note 1 based on imminent classification event promptly upon criteria.determining that time limit has been exceeded, or will likely be exceeded.79 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification HS7 Other conditions exist which in HS7 Other conditions existing that in the None the judgment of the Emergency judgment of the Site Emergency Director warrant declaration of a Coordinator warrant declaration of a Site Site Area Emergency.

Area Emergency MODE: All MODE: All NEI Ex. BNP EA E NEI Example EAL Wording EAL BNP EAL Wording Difference/Deviation Justification EAL # EAL #Other conditions exist which in HS7.1 Other conditions exist which in the None the judgment of the Emergency judgment of the Site Emergency Director indicate that events are Coordinator indicate that events are in in progress or have occurred progress or have occurred which involve which involve actual or likely actual or likely major failures of plant major failures of plant functions functions needed for protection of the needed for protection of the public or HOSTILE ACTION that results in public or HOSTILE ACTION that intentional damage or malicious acts, (1)results in intentional damage or toward site personnel or equipment that malicious acts, (1) toward site could lead to the likely failure of or, (2) that personnel or equipment that could prevent effective access to equipment lead to the likely failure of or, (2) needed for the protection of the public.that prevent effective access to Any releases are not expected to result in equipment needed for the exposure levels which exceed EPA protection of the public. Any Protective Action Guideline exposure releases are not expected to levels beyond the site boundary.result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.80 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification HG1 HOSTILE ACTION resulting in HG1 HOSTILE ACTION resulting in loss of None loss of physical control of the physical control of the facility facility. MODE: All MODE: All NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification

a. A HOSTILE ACTION is HG1.1 A HOSTILE ACTION is occurring or has The Security Shift Supervision is the site-specific security occurring or has occurred within occurred within the PROTECTED AREA shift supervision.

the PROTECTED AREA as ~as reported by the Security ShiftDeeethwod"nrl"aerracityastiseudn. the PROTECTED AREA as reported by the (site-specific Supervision Deleted the word "control" after "reactivity as it is redundant. security shift supervision). AND EITHER of the following has AND occurred: b. EITHER of the following has Any of the following safety functions occurred: cannot be controlled or maintained

1. ANY of the following safety 0 Reactivity functions cannot be a RPV water level controlled or maintained.

RCS heat removal* Reactivity control OR Core cooling[PWF]/RPV water Damage to spent fuel has occurred level [BWR] or is IMMINENT e RCS heat removal OR 2. Damage to spent fuel has occurred or is IMMINENT.81 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification HG7 Other conditions exist which in HG7 Other conditions exist which in the None the judgment of the Emergency judgment of the Site Emergency Director warrant declaration of a Coordinator warrant declaration of a General Emergency General Emergency MODE: All MODE: All NEI Ex. BNP EA E NEI Example EAL Wording EAL BNP EAL Wording Difference/Deviation Justification EAL # EAL #Other conditions exist which in HG7.1 Other conditions exist which in the None the judgment of the Emergency judgment of the Site Emergency Director indicate that events are Coordinator indicate that events are in in progress or have occurred progress or have occurred which involve which involve actual or actual or IMMINENT substantial core IMMINENT substantial core degradation or melting with potential for degradation or melting with loss of containment integrity or HOSTILE potential for loss of containment ACTION that results in an actual loss of integrity or HOSTILE ACTION physical control of the facility.

Releases that results in an actual loss of can be reasonably expected to exceed physical control of the facility. EPA Protective Action Guideline Releases can be reasonably exposure levels offsite for more than the expected to exceed EPA immediate site area.Protective Action Guideline exposure levels offsite for more than the immediate site area.82 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Category S System Malfunction 83 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification SUl Loss of all offsite AC power SUl Loss of all offsite AC power Deleted "Hot Standby" because BWRs do not have this operating capability to emergency buses for capability to emergency buses for mode.15 minutes or longer. 15 minutes or longer MODE: Power Operation, Startup, MODE: 1 -Power Operation, 2 -Hot Standby, Hot Shutdown Startup, 3 -Hot Shutdown NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 Loss of ALL offsite AC power SUI.1 Loss of all offsite AC power 4 KV essential buses El (E3) and E2(E4) are the site-specific capability to (site-specific capability to Emergency 4 KV emergency buses.emergency buses) for 15 minutes Buses El (E3) and E2(E4) for or longer. > 15 min. (Note 1)Note The Emergency Director should N/A Note 1: The SEC should declare The classification timeliness note has been standardized across the declare the Unusual Event the event promptly upon BNP EAL scheme by referencing the "time limit" specified within the promptly upon determining that determining that time EAL wording.15 minutes has been exceeded, limit has been or will likely be exceeded. exceeded, or will likely be exceeded.84 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification SU2 UNPLANNED loss of Control SU3 UNPLANNED loss of Control Deleted "Hot Standby' because BWRs do not have this operating Room indications for 15 minutes Room indications for 15 minutes mode.or longer, or longer.MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. BNP EA E NEI Example EAL Wording EAL BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 An UNPLANNED event results in SU3.1 An UNPLANNED event results in The site-specific Safety System Parameter list is tabulated in Table the inability to monitor one or the inability to monitor one or S-1.more of the following parameters more Table S-1 parameters from Torus water level and torus temperature are the BNP parameters from within the Control Room for within the Control Room for > 15 corresponding to suppression pool level and suppression pool 15 minutes or longer. min. (Note 1) temperature. Note The Emergency Director should N/A Note 1: The SEC should declare The classification timeliness note has been standardized across the declare the Unusual Event the event promptly upon BNP EAL scheme by referencing the "time limit" specified within the promptly upon determining that determining that time EAL wording.15 minutes has been exceeded, limit has been exceeded, or will likely be exceeded. or will likely be exceeded.85 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP[BWR parameter lisfl [PWR parameter lisfl Reactor Power Reactor Power RPV Water Level RCS Level RPV Pressure RCS Pressure Primary Containment Pressure In-Core/Core Exit Temperature Suppression Pool Level Levels in at least (site-specific number)steam generators Suppression Pool Temperature Steam Generator Auxiliary or Emergency Feed Water Flow Table S-1 Safety System Parameters

Reactor power* RPV water level* RPV pressure* Primary containment pressure* Torus water level* Torus temperature 86 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification SU3 Reactor coolant activity greater SU4 Reactor coolant activity greater Deleted "Hot Standby" because BWRs do not have this operating than Technical Specification than Technical Specification mode.allowable limits, allowable limits MODE: Power Operation, Startup, MODE: 1 -Power Operation, 2 -Hot Standby, Hot Shutdown Startup, 3 -Hot Shutdown NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification (Site-specific radiation monitor) SU4.1 Steam Jet Air Ejector Radiation The Steam Jet Air Ejector radiation monitor setpoint provides reading greater than (site-specific Monitor 1 (2)D1 2-RM-K601 A /B reasonable assurance that the total body exposure to an individual value). Hi-Hi alarm (Process Off-Gas at the exclusion area boundary will not exceed a small fraction of Rad Hi-Hi alarm 1 (2)UA-03 4-2) the limits of 1 OCFR1 00 in the event of an inadvertent release via> 15 min. (Note 1) the condenser air ejector. The 15 min. criteria is consistent with the off-gas timer duration isolating the off-gas system.2 Sample analysis indicates that a SU4.2 Coolant activity > 0.2 pCi/gm I- The specific iodine activity is limited to < 0.2 pCi/gm Dose reactor coolant activity value is 131 dose equivalent for > 48 Equivalent 1-131. This limit ensures the source term assumed in the greater than an allowable limit hours safety analysis for the Main Steam Line Break (MSLB) is not specified in Technical OR exceeded, so any release of radioactivity to the environment during Specifications.

Coolant activity > 4.0 pCi/gm I- an MSLB is less than a small fraction of the 10 CFR 50.67 limits.131 dose equivalent instantaneous The upper limit of 4.0 pCi/gm Dose Equivalent 1-131 ensures that the thyroid dose from an MSLB will not exceed the dose guidelines of 10 CFR 50.67 or Control Room operator dose limits specified in GDC 19 of 10 CFR 50, Appendix A.None N/A Note 1: The SEC should declare The classification timeliness note has been standardized across the the event promptly upon BNP EAL scheme by referencing the "time limit" specified within the determining that time limit has EAL s b r t m " f h been exceeded, or will likely be EAL wording.exceeded.87 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification SU4 RCS leakage for 15 minutes or SU5 RCS leakage for 15 minutes or Deleted "Hot Standby" because BWRs do not have this operating longer. longer mode.MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 RCS unidentified or pressure SU5.1 RCS unidentified or pressure Example EALs #1, 2 and 3 have been combined into a single EAL boundary leakage greater than boundary leakage > 10 gpm for > for usability.(site-specific value) for 15 15 min.minutes or longer. OR 2 RCS identified leakage greater RCS identified leakage > 25 gpm than (site-specific value) for 15 for > 15 min.minutes or longer. OR 3 Leakage from the RCS to a Leakage from the RCS to a location outside containment location outside Primary greater than 25 gpm for 15 Containment > 25 gpm for > 15 minutes or longer. min.(Note 1)Note The Emergency Director should N/A Note 1: The SEC should declare The classification timeliness note has been standardized across the declare the Unusual Event the event promptly upon BNP EAL scheme by referencing the "time limit" specified within the promptly upon determining that determining that time EAL wording.15 minutes has been exceeded, limit has been or will likely be exceeded. exceeded, or will likely be exceeded.88 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification SU5 Automatic or manual (trip SU6 Automatic or manual scram fails Included Mode 2 Startup consistent with developer note. Reactor[PWR] / scram [BWR]) fails to to shut down the reactor power can be above the APRM downscale shutdown threshold of 2%shutdown the reactor. MODE: 1 -Power Operation, 2 -while still in Mode 2.MODE: Power Operation Startup NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 a. An automatic (trip [PWR] / SU6.1 An automatic scram did not shut Added the phrase "... after any RPS setpoint is exceeded "to clarify scram [BWR]) did not shutdown down the reactor after any RPS that it is a failure of the automatic scram when a valid scram signal the reactor. setpoint is exceeded has been exceeded.AND AND Added "subsequent automatic scram or..." to take credit for automatic initiation of ARI subsequent to an RPS failure to complete b. A subsequent manual action A subsequent automatic scram automatic initiation of ARI to intae an taken at the reactor control or manual scram action taken at automatic scram Adegfature of ARI.consoles is successful in the reactor control console shutting down the reactor. (Manual PBs, Mode Switch, ARI) Reactor power below 2% is the site-specific indication of a is successful in shutting down the successful reactor scram.reactor as indicated by reactor Manual scram pushbuttons, Mode Switch and manual initiation of power < 2% (APRM downscale) ARI are the manual scram actions credited for this EAL as these (Note 8). actions can be quickly taken at the reactor control consoles.2 a. A manual trip ([PWR] / SU6.2 A manual scram did not shut Added the phrase "... after any manual scram action was initiated" to scram [BWR]) did not shutdown down the reactor after any clarify that it is a failure of any manual scram when an actual manual the reactor. manual scram action was scram signal has been inserted.AND initiated Combined conditions b.1 and b.2 into a single statement to simplify b. EITHER of the following: AND the presentation.

1. A subsequent manual A subsequent automatic scram Reactor power below 2% is the site-specific indication of a or manual scram action taken at successful reactor scram.ction tknathe rc the reactor control console Manual scram pushbuttons, Mode Switch and manual initiation of successful in shutting (Manual PBs, Mode Switch, ARI) ARI are the manual scram actions credited for this EAL as these down the reactort is successful in shutting down the actions can be quickly taken at the reactor control consoles.down the reactor. ~reactor as indicated by reactor ______________________________

89 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP OR power < 2% (APRM downscale) 2 A subsequent automatic (Note 8)(trip [PWR] / scram[BWR]) is successful in shutting down the reactor.Notes Note: A manual action is any N/A Note 8: A manual scram action Added the scram to actions to be consistent with EAL wording.operator action, or set of actions, is any operator action, or which causes the control rods to set of actions, which be rapidly inserted into the core, causes the control rods and does not include manually to be rapidly inserted driving in control rods or into the core, and does implementation of boron not include manually injection strategies. driving in control rods or implementation of boron injection strategies. N/A 90 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification SU6 Loss of all onsite or offsite SU7 Loss of all onsite or offsite Deleted "Hot Standby" because BWRs do not have this operating communications capabilities, communications capabilities, mode.MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 Loss of ALL of the following SU7.1 Loss of all Table S-3 onsite Example EALs #1, 2 and 3 have been combined into a single EAL onsite communication methods: communication methods for simplification of presentation.(site-specific list of OR Table S-2 provides a site-specific list of onsite, (offsite) ORO and communications methods) Loss of all Table S-3 offsite NRC communications methods.2 Loss of ALL of the following communication methods ORO communications methods: OR (site-specific list of Loss of all Table S-3 NRC communications methods) communication methods 3 Loss of ALL of the following NRC communications methods: (site-specific list of communications methods)91 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP Table S-3 Communication Methods System Onsite Offsite NRC Public Address System X PBX Telephone System X X X Corporate Telephone System X X X Commercial Telephones X X X Satellite Phones X X Cellular Phones X X NRC emergency Telecommunications X System 92 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification SU7 Failure to isolate containment or N/A N/A This IC and its associated example EALs are applicable to PWRs loss of containment pressure only and therefore not included.control. [PWR]MODE: Hot Standby, Hot Shutdown NEI Ex. NEI Example EAL Wording BNP BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 a. Failure of containment to N/A N/A This IC and its associated example EALs are applicable to PWRs isolate when required by an only and therefore not included.actuation signal.AND b. ALL required penetrations are not closed within 15 minutes of the actuation signal.2 a. Containment pressure N/A N/A This IC and its associated example EALs are applicable to PWRs greater than (site-specific only and therefore not included.pressure). AND b. Less than one full train of (site-specific system or equipment) is operating per design for 15 minutes or longer.93 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification SAl Loss of all but one AC power SA1 Loss of all but one AC power Deleted "Hot Standby" because BWRs do not have this operating source to emergency buses for source to emergency buses for mode.15 minutes or longer. 15 minutes or longer.MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. BNP EA E NEI Example EAL Wording EAL BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 a. AC power capability to (site- SA1.1 AC power capability to 4 KV emergency buses El (E3) and E2(E4) are the site-specific specific emergency buses) is Emergency 4 KV Buses El (E3) emergency buses.reduced to a single power and E2(E4) reduced to a single source for 15 minutes or longer, power source for > 15 min.(Note 1) Because 2 RHR pumps on each unit are powered from the AND unaffected unit, the words "unit-specific" have been added to clarify b. Any additional single power AND that the cross-connected RHR pump power cannot be credited as source failure will result in a loss Any additional single power an AC power source relative to this EAL.of all AC power to SAFETY source failure will result in loss of SYSTEMS. all unit-specific AC power to SAFETY SYSTEMS Note The Emergency Director should N/A Note 1: The SEC should The classification timeliness note has been standardized across declare the Alert promptly upon declare the event the BNP EAL scheme by referencing the "time limit" specified determining that 15 minutes has promptly upon within the EAL wording.been exceeded, or will likely be determining that time exceeded. limit has been exceeded, or will likely be exceeded.94 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification SA2 UNPLANNED loss of Control SA3 UNPLANNED loss of Control Deleted "Hot Standby" because BWRs do not have this operating Room indications for 15 minutes Room indications for 15 minutes mode.or longer with a significant or longer with a significant transient in progress. transient in progress.MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEl Ex. BNP EA E NEI Example EAL Wording EAL BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 An UNPLANNED event results in SA3.1 An UNPLANNED event results in The site-specific Safety System Parameters are listed in Table S-1.the inability to monitor one or the inability to monitor one or Torus water level and torus temperature are the BNP parameters more of the following parameters more Table S-1 parameters from coruspwater level and supre esBn pool from within the Control Room for within the Control Room for 2 ! corresponding to suppression pool level and suppression pool 15 minutes or longer. min. (Note 1) temperature. AND AND The significant transient list has been tabularized in Table S-2 for ease of use.ANY of the following transient Any significant transient is in events in progress. progress, Table S-2* Automatic or manual runback greater than 25%thermal reactor power* Electrical load rejection greater than 25% full electrical load" Reactor scram [BWR] / trip[PWR]* ECCS (SI) actuation" Thermal power oscillations 95 of 107 EAL Comparison Matrix OSSE Project #13-0703 BNP greater than 10% [BWR]Note The Emergency Director should N/A Note 1: The SEC should declare The classification timeliness note has been standardized across the declare the Unusual Event the event promptly upon BNP EAL scheme by referencing the "time limit" specified within the promptly upon determining that determining that time EAL wording.15 minutes has been exceeded, limit has been exceeded, or will likely be exceeded. or will likely be exceeded.96 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP[BWR parameter list] [PWR parameter list]Reactor Power Reactor Power RPV Water Level RCS Level RPV Pressure RCS Pressure Primary Containment Pressure In-Core/Core Exit Temperature Suppression Pool Level Levels in at least (site-specific number)steam generators Suppression Pool Temperature Steam Generator Auxiliary or Emergency Feed Water Flow Table S-1 Safety System Parameters

Reactor power* RPV water level* RPV pressure* Primary containment pressure* Torus water level* Torus temperature Table S-2 Significant Transients
Reactor scram* Runback > 25% rated thermal power* Electrical load rejection

> 25% electrical load* ECCS injection* Thermal power oscillations > 10% (peak to peak)97 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification SA5 Automatic or manual (trip [PWR] SA6 Automatic or manual scram fails Included Mode 2 Startup consistent with developer note. Reactor/ scram [BWR]) fails to shutdown to shut down the reactor and power can be above the APRM downscale shutdown threshold of the reactor, and subsequent subsequent manual actions 2% while still in Mode 2.manual actions taken at the taken at the reactor control reactor control consoles are not consoles are not successful in successful in shutting down the shutting down the reactor reactor. MODE: 1 -Power Operation, 2 -MODE: Power Operation Startup NEI Ex. NEI Example EAL Wording BNP BNP EAL Wording Difference/Deviation Justification EAL # EAL #a. An automatic or manual (trip SA6.1 An automatic or manual scram Reactor power below 2% is the site-specific indication of a[PWR] / scram [BWR]) did not fails to shut down the reactor successful reactor scram.shutdown the reactor. AND Manual scram pushbuttons, Mode Switch and manual initiation of AND Manual scram actions taken at ARI are the manual scram actions credited for this EAL as these Manual scram actions taken at tactions can be quickly taken at the reactor control consoles.b. Manual actions taken at the the reactor control console reactor control consoles are not (Manual PBs, Mode Switch, successful in shutting down the ARI) are not successful in reactor. shutting down the reactor as indicated by reactor power > 2%(Note 8)Notes Note: A manual action is any N/A Note 8: A manual scram action Added the word "scram" to actions consistent with EAL wording.operator action, or set of actions, is any operator action, or which causes the control rods to set of actions, which be rapidly inserted into the core, causes the control rods and does not include manually to be rapidly inserted driving in control rods or into the core, and does implementation of boron injection not include manually strategies. driving in control rods or implementation of boron 98 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP injection strategies. 99 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification Hazardous event affecting a SA8 Hazardous event affecting a Deleted "Hot Standby" because BWRs do not have this operating SAFETY SYSTEM needed for SAFETY SYSTEM required for mode.the current operating mode. the current operating mode.MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. BNP EAL # NEI Example EAL Wording EAL # BNP EAL Wording Difference/Deviation Justification

a. The occurrence of ANY of SA8.1 The occurrence of any Table S- The hazardous events have been listed in Table S-4 to improve the the following hazardous events: 4 hazardous event readability of the BNP EAL.o Seismic event (earthquake)

AND The NEI list of hazardous events includes all BNP hazardous events." Internal or external flooding EITHER of the following: No additional hazardous events could be identified. event Event damage has caused* High winds or tornado strike indications of degraded" FIRE performance in at least one train of a SAFETY SYSTEM" EXPLOSION required for the current* (site-specific hazards) operating mode" The event has caused h Other events with similar VISIBLE DAMAGE to a hazard characteristics as SAFETY SYSTEM determined by the Shift component or structure Manager required for the current AND operating mode b. EITHER of the following:

1. Event damage has caused indications of degraded performance in at least one train of a SAFETY 100 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP SYSTEM needed for the SYSTEM needed for the current operating mode.OR 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.Table S-4 Hazardous Events* Seismic event (earthquake)

Internal or external FLOODING event* High winds or tornado strike* FIRE* EXPLOSION* Other events with similar hazard characteristics as determined by the Shift Manager 101 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification SS1 Loss of all offsite and all onsite SS1 Loss of all offsite and all onsite Deleted "Hot Standby" because BWRs do not have this operating AC power to emergency buses AC power to emergency buses mode.for 15 minutes or longer, for 15 minutes or longer MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. BNP EAL # NEI ExampEAL AL Wording EAL # BNP EAL Wording Difference/Deviation Justification 1 Loss of ALL offsite and ALL SS1.1 Loss of all offsite and all onsite 4 KV essential buses El (E3) and E2(E4) are the site-specific onsite AC power to (site-specific AC power capability to emergency buses.emergency buses) for 15 Emergency 4 KV Buses El (E3)minutes or longer, and E2(E4) for > 15 min. (Note 1)Note The Emergency Director should N/A Note 1: The SEC should The classification timeliness note has been standardized across the declare the Unusual Event declare the event BNP EAL scheme by referencing the "time limit" specified within promptly upon determining that promptly upon the EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded.

limit has been exceeded, or will likely be exceeded.102 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification SS5 Inability to shutdown the reactor SS6 Inability to shut down the Included Mode 2 Startup consistent with developer note. Reactor causing a challenge to (core reactor causing a challenge to power can be above the APRM downscale shutdown threshold of cooling [PWF] / RPV water level RPV water level or RCS heat 2% while still in Mode 2.[BWR]) or RCS heat removal, removal MODE: Power Operation MODE: 1 -Power Operation, 2 -Startup NEI Ex. NEI Example EAL Wording EALBNP BNP EAL Wording Difference/Deviation Justification

a. An automatic or manual (trip SS6.1 An automatic or manual scram Reactor power < 2% is the site-specific indication of a successful

[PWR] / scram [BWR]) did not fails to shut down the reactor reactor scram.shutdown the reactor. AND Deleted the term 'manual actions" from the second conditional. For AND All actions to shut down the generic IC SS5, all actions to shutdown the reactor can be credited, b. All manual actions to reactor are not successful as including emergency boration which is not considered a "manual" shutdown the reactor have been indicated by reactor power > 2% scram action.unsuccessful. AND EITHER Indication of an inability to adequately remove heat from the core AND occurs when RPV water level cannot be restored and maintained 9 RPV level cannot be above LL-4, which is the EOP RPV water level indicative of a loss of c. EITHER of the following restored and maintained adequate core cooling.conditions exist: > LL-4 or cannot be Indication of an inability to adequately remove heat from the RCS (Site-specific indication of determined occurs when torus water temperature cannot be maintained below an inability to adequately e Suppression pool water the Heat Capacity Temperature Limit.remove heat from the temperature and RPV core) pressure cannot be (Site-specific indication of maintained below the HCTL an inability to adequately remove heat from the RCS)103 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification SS8 Loss of all Vital DC power for 15 SS2 Loss of all vital DC power for 15 Deleted "Hot Standby' because BWRs do not have this operating minutes or longer. minutes or longer, mode.MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording BNP #EAL # EAL BNP EAL Wording Difference/Deviation Justification 1 Indicated voltage is less than SS2.1 Loss of all 125 VDC power 105 VDC is the site-specific minimum vital DC bus design voltage (site-specific bus voltage value) based on battery bus voltage per Technical Specification 3.8.4.on ALL (site-specific Vital DC indications < 105 VDC on all 1(2)A-1, A-2, B-1 and B-2 are the site-specific vital DC buses.busses) for 15 minutes or longer, vital DC buses 1 (2)A-1, A-2, B-1 and B-2 for > 15 min. (Note 1)Note The Emergency Director should N/A Note 1: The SEC should The classification timeliness note has been standardized across the declare the Unusual Event declare the event promptly upon BNP EAL scheme by referencing the "time limit" specified within the promptly upon determining that determining that time limit has EAL wording.15 minutes has been exceeded, been exceeded, or will likely be or will likely be exceeded. exceeded.104 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BN P NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification SG1 Prolonged loss of all offsite and SG1 Prolonged loss of all offsite and Combined NEI ICs SG1 and SG8 under the loss of power category all onsite AC power to all onsite AC power to for usability. emergency buses. emergency buses OR loss of all Deleted "Hot Standby" because BWRs do not have this operating MODE: Power Operation, emergency AC and vital DC mode.Startup, Hot Standby, Hot power sources for 15 minutes or Shutdown longer MODE: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown NEI Ex. BNP EA NEI Example EAL Wording EAL BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 a. Loss of ALL offsite and ALL SG1.1 Loss of all offsite and all onsite 4 KV emergency buses El (E3) and E2(E4) are the site-specific onsite AC power to (site-specific AC power capability to emergency buses.emergency buses). Emergency 4 KV Buses El (E3) 4 hours is the site-specific SBO coping analysis time.AND and E2(E4) Indication of an inability to adequately remove heat from the core b. EITHER of the following: AND EITHER occurs when RPV water level cannot be restored and maintained

Restoration of at least one above LL-4 which is the EOP RPV water level indicative of a loss of Restoration of at least emergency bus in < 4 hours adequate core cooling.one AC emergency bus in emer y bus in less than (site-specific is not likely (Note 1)hours) is not likely. e RPV water level cannot be restored and maintained

>(Site-specific indication of MSCRWL (LL-4)an inability to adequately remove heat from the core)Note The Emergency Director should N/A Note 1: The SEC should The classification timeliness note has been standardized across the declare the General Emergency declare the event BNP EAL scheme by referencing the "time limit" specified within the promptly upon determining that promptly upon EAL wording.(site-specific hours) has been determining that time exceeded, or will likely be limit has been exceeded. exceeded, or will likely be exceeded.105 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP BNP NEI IC# NEI IC Wording IC#(s) BNP IC Wording Difference/Deviation Justification SG8 Loss of all AC and Vital DC SG1 Prolonged loss of all offsite and Combined NEI ICs SG1 and SG8 under the loss of power category power sources for 15 minutes or all onsite AC power to for usability. longer. emergency buses OR loss of all Deleted "Hot Standby" because BWRs do not have this operating MODE: Power Operation, emergency AC and vital DC mode.Startup, Hot Standby, Hot power sources for 15 minutes or Shutdown longer.MODE: 1 -Power Operation, 2-Startup, 3 -Hot Shutdown NEI Ex. BNP EA E NEI Example EAL Wording EAL BNP EAL Wording Difference/Deviation Justification EAL # EAL #1 a. Loss of ALL offsite and ALL SG1.2 Loss of all offsite and all 4 KV essential buses El (E3) and E2(E4) are the site-specific onsite AC power to (site-specific onsite AC power capability to emergency buses.emergency buses) for 15 Emergency 4 KV Buses 105 VDC is the site-specific minimum vital DC bus design voltage minutes or longer. El (E3) and E2(E4) for > 15 per Technical Specification 3.8.4.AND AinD DC buses 1(2)A-1, A-2, B-1 and B-2 are the site-specific vital DC b. Indicated voltage is less than AND buses.(site-specific bus voltage value) Loss of all 125 VDC power on ALL (site-specific Vital DC based on battery bus voltage busses) for 15 minutes or longer, indications < 105 VDC on all vital DC buses 1(2)A-1, A-2, B-1 and B-2 for -> 15 min.(Note 1)Note The Emergency Director should N/A Note 1: The SEC should The classification timeliness note has been standardized across the declare the Unusual Event declare the event BNP EAL scheme by referencing the "time limit" specified within the promptly upon determining that promptly upon EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded. limit has been exceeded, or will likely 106 of 107 EAL Comparison Matrix OSSI Project #13-0703 BNP be exceeded.107 of 107}}

ML15044A245

Text

Subject:

Enclosures:

Subject:

2.1. Proposed

4.1. Applicable

4.0 REFERENCES

1.0 PURPOSE

2.0 DISCUSSION

2.1 Background

2.2 Fission

2.3 Fission

2.5 Technical

3.0 GUIDANCE

3.1 General

3.1.1 Classification

3.1.3 Imminent

3.2.8 Retraction

4.0 REFERENCES

4.1 Developmental

5.1 Definitions

7.0 ATTACHMENTS

7.1 Attachment

SUMMARY

4.0 REFERENCES

1.0 PURPOSE

2.0 DISCUSSION

2.1 Background

2.2 Fission

2.3 Fission

2.5 Technical

3.0 GUIDANCE

3.1 General

3.1.1 Classification

3.1.3 Imminent

3.2.8 Retraction

4.0 REFERENCES

4.1 Developmental

5.1 Definitions

7.0 ATTACHMENTS

SUMMARY

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