ML15344A153

Issuance of Amendments Regarding Emergency Action Level Scheme Upgrade (CAC Nos. MF5766 and MF5767)

ML15344A153
Person / Time
Site:	Brunswick
Issue date:	01/08/2016
From:	Andrew Hon Plant Licensing Branch II
To:	William Gideon Progress Energy Carolinas
Hon A
References
CAC MF5766, CAC MF5767
Download: ML15344A153 (31)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. William R. Gideon Site Vice President Brunswick Steam Electric Plant 8470 River Road, SE M/C BNP001 Southport, NC 28461 January 8, 2016

SUBJECT:

BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 -ISSUANCE OF AMENDMENTS REGARDING EMERGENCY ACTION LEVEL SCHEME UPGRADE (CAC NOS. MF5766 AND MF5767)

Dear Mr. Gideon:

The Commission has issued the enclosed Amendment Nos. 268 and 296 to Renewed Facility Operating License Nos. DPR-71 and DPR-62, respectively, for the Brunswick Steam Electric Plant, Units 1 and 2 (BSEP). These amendments are in response to your application dated January 30, 2015, as supplemented by letter dated November 23, 2015. Duke Energy Progress, Inc. (Duke Energy, the licensee) requested a change to the emergency plan for BSEP. The change revises the emergency action level scheme for each unit based on the Nuclear Energy Institute (NEI) document NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," dated November 2012. NEI 99-01, Revision 6, was endorsed by the U.S. Nuclear Regulatory Commission (NRC) by letter dated March 28, 2013. The NRC staff has completed its review of the above information provided by the licensee and approved the request based on the enclosed Safety Evaluation (SE). The NRC staff has determined that its documented SE does not contain sensitive security-related information pursuant to Title 10 of the Code of Federal Regulations (CFR), Section 2.390, "Public inspections, exemptions, requests for withholding." However, the NRC will delay placing the enclosed SE in the public document room for a period of 10 working days from the date of this letter to provide Duke Energy with the opportunity to comment on any sensitive aspects. If you believe that any information in the SE contains sensitive information, please identify such information line-by-line and define the basis pursuant to the criteria of 10 CFR 2.390. If you do not identify sensitive information after 10 working days, the enclosed SE will be made publicly available.

W. Gideon A Notice of Issuance will be included in the Commission's Biweekly Federal Register Notice. Docket Nos. 50-324 and 50-325

Enclosures:

1. Amendment No. 268 to DPR-71 2. Amendment No. 296 to DPR-62 3. Safety Evaluation cc w/enclosures:

Distribution via Listserv Sincerely, Andrew Hon, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY PROGRESS.

INC. DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 268 Renewed License No. DPR-71 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Duke Energy Progress, Inc., dated January 30, 2015, as supplemented by letter dated November 23, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-71 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 268, are hereby incorporated in the license. Duke Energy Progress, Inc. shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 180 days from the date of issuance.

Attachment:

Changes to the Renewed Facility Operating License Date of Issuance:

January 8, 2016 FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ATTACHMENT TO LICENSE AMENDMENT NO. 268 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 Replace the following page of Renewed Facility Operating License No. DPR-71 with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change. REMOVE Page 6 INSERT Page 6 (c) Transition License Conditions

1. Before achieving full compliance with 10 CFR 50.48(c), as specified by 2. below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2. above. 2. The licensee shall implement the modifications to its facility, as described in Table S-1, "Plant Modifications Committed," of Duke letter BSEP 14-0122, dated November 20, 2014, to complete the transition to full compliance with 10 CFR 50.48( c) by the startup of the second refueling outage for each unit after issuance of the safety evaluation.

The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

3. The licensee shall complete all implementation items, except item 9, listed in LAR Attachment S, Table S-2, "Implementation Items," of Duke letter BSEP 14-0122, dated November 20, 2014, within 180 days after NRC approval unless the 180th day falls within an outage window; then, in that case, completion of the implementation items, except item 9, shall occur no later than 60 days after startup from that particular outage. The licensee shall complete implementation of LAR Attachment S, Table S-2, Item 9, within 180 days after the startup of the second refueling outage for each unit after issuance of the safety evaluation.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions hereafter in effect; and is subject to the additional conditions specified or incorporated below: (1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2923 megawatts thermal. (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 268, are hereby incorporated in the license. Duke Energy Progress, Inc. shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 203 to Renewed Facility Operating License DPR-71, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 203. For SRs that existed prior to Amendment 203, including SRs with modified acceptance criteria and SRs whose frequency of Renewed License No. DPR-71 Amendment No. 268 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY PROGRESS.

INC. DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 296 Renewed License No. DPR-62 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Duke Energy Progress, Inc., dated January 30, 2015, as supplemented by letter dated November 23, 2015,, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 2 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-62 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 296, are hereby incorporated in the license. Duke Energy Progress, Inc. shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 180 days from the date of issuance.

Attachment:

Changes to the Renewed Facility Operating License Date of Issuance:

January 8, 2016 FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ATTACHMENT TO LICENSE AMENDMENT NO. 296 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-62 DOCKET NO. 50-324 Replace the following page of Renewed Facility Operating License No. DPR-62 with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change. REMOVE Page 6 INSERT Page 6 (c) Transition License Conditions

1. Before achieving full compliance with 10 CFR 50.48(c), as specified by 2. below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2. above. 2. The licensee shall implement the modifications to its facility, as described in Table S-1, "Plant Modifications Committed," of Duke letter BSEP 14-0122, dated November 20, 2014, to complete the transition to full compliance with 10 CFR 50.48(c) by the startup of the second refueling outage for each unit after issuance of the safety evaluation.

The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

3. The licensee shall complete all implementation items, except Item 9, listed in LAR Attachment S, Table S-2, "Implementation Items," of Duke letter BSEP 14-0122, dated November 20, 2014, within 180 days after NRC approval unless the 180th day falls within an outage window; then, in that case, completion of the implementation items, except item 9, shall occur no later than 60 days after startup from that particular outage. The licensee shall complete implementation of LAR Attachment S, Table S-2, Item 9, within 180 days after the startup of the second refueling outage for each unit after issuance of the safety evaluation.

B. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: (1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2923 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 296, are hereby incorporated in the license. Duke Energy Progress, Inc. shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 233 to Renewed Facility Operating License DPR-62, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 233. For SRs that existed prior to Amendment 233, Renewed License No. DPR-62 Amendment No. 296 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 268 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-71 AND AMENDMENT NO. 296 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-62 DUKE ENERGY PROGRESS.

INC. BRUNSWICK STEAM ELECTRIC PLANT. UNITS 1 AND 2 DOCKET NOS. 50-325 and 50-324

1.0 INTRODUCTION

By application dated January 30, 2015, as supplemented by letter dated November 23, 2015 (Agencywide Documents Access and Management System (ADAMS) Package Accession Nos. ML15044A198 and ML15350A105, respectively), Duke Energy Progress, Inc. (Duke Energy, the licensee) requested a change to the emergency plan for the Brunswick Steam Electric Plant, Units 1 and 2 (BSEP). The proposed change revises the emergency action level (EAL) scheme for each unit based on the Nuclear Energy Institute (NEI) document NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors,

dated November 2012. NEI 99-01, Revision 6, was endorsed by the U.S. Nuclear Regulatory Commission (NRC or Commission) by letter dated March 28, 2013. The supplement dated November 23, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on April 28, 2015 (80 FR 23602).

2.0 REGULATORY EVALUATION

The applicable regulations and guidance for the emergency plans follow. 2.1 Regulations Title 1 O of the Code of Federal Regulations (10 CFR), Section 50.47, "Emergency plans," sets forth emergency plan requirements for nuclear power plant facilities.

The regulation in 1 O CFR 50.47(a)(1

)(i) states, in part: ( ... ] no initial 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 Enclosure 3 protective measures can and will be taken in the event of a radiological emergency.

Section 50.47(b) establishes the standards that the onsite and offsite emergency response plans must meet for NRC staff to make a positive finding that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.

Planning standard (4) of this section requires that onsite and offsite emergency response plans meet the following standard:

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

Section 50.47(b)(4) emphasizes use of a standard emergency classification and action level scheme, ensuring that implementation methods are relatively consistent throughout the industry for a given reactor and containment design, while simultaneously providing an opportunity for a licensee to modify its EAL scheme as necessary to address plant-specific design considerations or preferences.

Section IV.B of Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities," to 1 O CFR Part 50, states, in part: The means to be used for determining the magnitude of, and for continually assessing the impact of, the release of radioactive materials shall be described, including emergency action levels that are to be used as criteria for determining the need for notification and participation of local and State agencies, the Commission, and other Federal agencies, and the emergency action levels that are to be used for determining when and what type of protective measures should be considered within and outside the site boundary to protect health and safety. The emergency action levels shall be based on in-plant conditions and instrumentation in addition to onsite and offsite monitoring.

By June 20, 2012, for nuclear power reactor licensees, these action levels must include hostile action that may adversely affect the nuclear power plant. 2.2 Guidance The EAL development guidance was initially established in Generic Letter (GL) 79-50 and was subsequently established in NUREG-0654/FEMA-REP-1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," which was endorsed as an approach for the development of an EAL scheme via NRC Regulatory Guide (RG) 1.101, Revision 2, "Emergency Planning and Preparedness for Nuclear Power Reactors." As industry and regulatory experience was gained with the implementation and use of EAL schemes, the industry issued revised EAL scheme development guidance to reflect lessons learned. To date, NUMARC/NESP-007 and NEI 99-01, Revisions 4, 5, and 6, were provided to the NRC for review and endorsement as generic (non-plant-specific)

EAL development guidance.

RG 1.101, Revisions 3 and 4, endorsed NUMARC/NESP-007 and NEI 99-01, Revision 4, as acceptable alternatives for licensees to consider in the development of their plant-specific EAL schemes and allowed licensees to develop plant-specific EALs based upon an alternative approach not endorsed by the NRC. NEI 99-01, Revision 5, was endorsed by the NRC as generic (non-plant-specific)

EAL scheme development guidance via letter dated February 22, 2008. NEI 99-01, Revision 6, dated November 2012 (ADAMS Accession No. ML12326A805), was endorsed by the NRC as generic (non-plant-specific)

EAL scheme development guidance via letter dated March 28, 2013 (ADAMS Accession No. ML12346A463).

The EAL development guidance contained in GL 79-50; NUREG-0654/FEMA-REP-1; NUMARC/NESP-007; and NEI 99-01, Revisions 4, 5, and 6, are all considered generic EAL scheme development guidance as they are not plant-specific and may not be entirely applicable for some reactor designs. However, the guidance contained in these documents bounds the most typical accident/event scenarios for which emergency response is necessary in a format that allows for industry standardization and consistent regulatory oversight.

Most licensees choose to develop plant-specific EAL schemes using the latest endorsed EAL development guidance with appropriate plant-specific alterations as applicable.

Pursuant to 1 O CFR Part 50, Appendix E,Section IV.B (2), a revision to an EAL must be approved by the NRC before implementation, if the licensee is changing from one EAL scheme to another EAL scheme. In summary, the NRC staff considers the following methods acceptable tor use in developing plant-specific EALs that meet the requirements of Section IV of Appendix E to 1 O CFR Part 50 and 10 CFR 50.47(b)(4), with the understanding that licensees may want to develop EALs that differ from the applicable guidance document as allowed in RG 1.101 and in the following applicable endorsement letters:

• Appendix 1, "Emergency Action Level Guidelines for Nuclear Power Plants," to NUREG-0654/FEMA-REP-1, dated November 1980;

• NUMARC/NESP-007, Revision 2, "Methodology tor Development of Emergency Action Levels," dated January 1992;

• NEI 99-01, Revision 4, "Methodology tor Development of Emergency Action Levels," dated January 2003;

• NEI 99-01, Revision 5, "Methodology for Development of Emergency Action Levels," dated February 2008; and

• NEI 99-01, Revision 6, "Development of Emergency Action Levels tor Non-Passive Reactors," dated November 2012. NRC Regulatory Issue Summary (RIS) 2003-18, with Supplements 1and2, "Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels," also provides guidance for developing or changing a standard emergency classification and action level scheme. In addition, this RIS and its supplements provide recommendations to assist licensees, consistent with Section IV.B of Appendix E to Part 50, in determining whether to seek prior NRC approval of deviations from the guidance.

Regardless of the generic EAL scheme development guidance document used by a licensee to develop its EAL scheme, or if a licensee chose to develop its EAL scheme using an alternative approach not endorsed by the NRC, or a combination of the two (most typical), the NRC staff reviews the EAL scheme to ensure it meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4).

3.0 TECHNICAL EVALUATION In its application, the licensee proposes to revise its current EAL scheme based on NEI 99-01, Revision 5, to one based on NEI 99-01, Revision 6. In its application and supplemental letter, the licensee submitted the proposed EAL scheme, the technical basis containing an evaluation and rationale for each proposed EAL change, and a comparison matrix providing a line-by-line comparison of the proposed BSEP Initiating Conditions, Mode Applicability, and EAL wording to that found in NEI 99-01, Revision 6. The comparison matrix also included a description of global changes applicable to the BSEP EAL scheme and a justification for any differences or deviations from NEI 99-01, Revision 6. The application states that the licensee used the terms "difference" and "deviation" as defined in RIS 2003-18, as supplemented, when comparing its proposed plant-specific EALs to the generic EALs in NEI 99-01, Revision 6. The NRC staff reviewed the proposed site-specific EAL scheme, technical basis, comparison matrix, and all additional information provided in the licensee's application and supplemental letter. The NRC staff found that both the current and proposed EALs have modifications from NEI 99-01, Revision 6, guidance, due to specific plant designs and licensee preference.

Although the EALs must be plant-specific, the NRC staff reviewed the proposed EALs for the following key characteristics of an effective EAL scheme to ensure consistency and regulatory stability:

• Consistency, including standardization of intent, if not in actual wording (i.e., the EALs would lead to similar decisions under similar circumstances at different plants);

• Human factors engineering and user friendliness;

• Potential for emergency classification level upgrade only when there is an increasing threat to public health and safety;

• Ease of upgrading and downgrading the emergency classification level;

• Thoroughness in addressing and disposing of the issues of completeness and accuracy raised regarding Appendix 1 to NUREG-0654 (i.e., the EALs are unambiguous and are based on site-specific indicators);

• Technical completeness for each classification level;

• Logical progression in classification for multiple events; and

• Objective and observable values. Based on its review, the NRC staff determined that the proposed EAL modifications do not alter the intent of any specific EAL described in NEI 99-01, Revision 6. The licensee chose to modify its proposed EAL scheme from the generic EAL scheme development guidance provided in NEI 99-01, Revision 6, in order to adopt a format that is better aligned with how it currently implements its EALs, as well as with plant-specific writer's guides and preferences.

The NRC staff determined that the proposed EAL scheme uses objective and observable values, is worded in a manner that addresses human factors engineering and user friendliness concerns, follows logical progressions for escalating events, and allows for event downgrading and upgrading based upon the potential risk to the public health and safety. Risk assessments were appropriately used to set the boundaries of the emergency classification levels and ensure that all EALs that trigger an emergency classification are in the same range of relative risk. In addition, the NRC staff has determined that the proposed EAL scheme is technically complete and consistent with EAL schemes implemented at similarly designed plants. Details regarding the NRC staff's review of specific EALs are provided below. To aid in understanding the nomenclature used in this safety evaluation, the following conventions are used:

• The first letter signifies the EAL category;

• The second letter signifies the emergency classification level: o G =General Emergency (GE), o S = Site Area Emergency (SAE), o A = Alert, and o U = Notification of Unusual Event (UE)

• The number denotes the sequential subcategory designation from the plant-specific EAL scheme. In addition, a set refers to all emergency classification levels (GE, SAE, A, and U) that share the same EAL category and subcategory.

This safety evaluation uses the numbering system from the plant-specific EAL scheme; however, the numbering system from the generic EAL scheme development guidance contained in NEI 99-01, Revision 6, is annotated in [brackets]

to aid in cross-referencing the site-specific EAL numbering convention with that of the guidance.

3.1 Category 'R' -Abnormal Radiological Release/Radiological Effluent 3.1.1 EAL Set RG1/RS1/RA1/RU1

[AG1/AS1/AA1/AU1]

This EAL set (or subgroup of the category) is based upon plant-specific indications of a release of radioactivity (gaseous and/or liquid). The NRC staff reviewed the licensee's evaluation and justification for plant-specific EAL changes associated with this set and has determined that the progression from UE to GE is appropriate and consistent with EAL scheme development guidance.

The licensee chose to modify this EAL set by using a site-specific implementation method that uses a modified numbering format and EAL sequence other than that provided in the generic EAL scheme development guidance.

However, based on its review, the NRC staff has determined that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance and with the site-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The instrumentation and set points derived for this EAL set are consistent with the overall EAL scheme development guidance, address the plant-specific implementation strategies provided, and are consistent with a standard EAL scheme, as required by 10 CFR 50.47(b)(4).

The NRC staff concludes that the plant-specific implementation method for this EAL set is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 1 O CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.1.2 EAL Set RG2/RS2/RA2/RU2

[AG2/AS2/AA2/AU2]

This EAL set is based upon plant-specific indications of fuel uncovery, including spent fuel stored in the spent fuel pool or refueling pathway. The NRC staff reviewed the licensee's evaluation and justification for plant-specific EAL changes associated with this set and has determined that the progression from UE to GE is appropriate and consistent with EAL scheme development guidance.

The SAE and GE classification levels for this specific accident progression are also bounded by indications available in the fission product barrier matrix, as well as in EA Ls RS 1 and RG 1. The licensee chose to modify this EAL set by using a site-specific implementation method that uses a modified numbering format and EAL sequence other than that provided in the generic EAL scheme development guidance.

However, based on its review, the NRC staff has determined that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance and with the site-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The instrumentation and set points derived for this EAL set are consistent with the overall EAL scheme development guidance, address the plant-specific implementation strategies provided, and are consistent with a standard EAL scheme, as required by 1 O CFR 50.47(b)(4).

The NRC staff concludes that the plant-specific implementation method for this EAL set is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b}(4), and is, therefore, acceptable for implementation.

3.1.3 EAL RA3 [AA3] This EAL is based upon radiation levels in the plant that limit normal access. This Alert EAL is primarily intended to ensure that the plant emergency response organization is activated to support the control room in removing the impediment to normal access, as well as assisting in quantifying potential damage to the fuel. Indications of increasing radiation levels in the plant are bounded by indication of fission product barrier loss or potential loss, as well as in RS1 and RG1. The licensee chose to modify this EAL by using a site-specific implementation method that uses a modified numbering format and EAL sequence other than that provided in the generic EAL scheme development guidance.

However, based on its review, the NRC staff has determined that the numbering, sequencing, and format of this EAL is consistent with the overall EAL scheme development guidance and with the site-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The NRC staff concludes that the plant-specific implementation method for this EAL is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.2 Category 'C' -Cold Shutdown/Refueling System Malt unction 3.2.1 EAL Set CG1/CS1/CA1/CU1

[CG1/CS1/CA1/CU1]

This EAL set is based upon a loss of reactor pressure vessel inventory and/or reactor coolant system (RCS) leakage. The NRC staff reviewed the licensee's evaluation and justification for plant-specific EAL changes associated with this set and has determined that the progression from UE to GE is appropriate and consistent with EAL scheme development guidance.

The NRC staff's review revealed that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance and with the plant-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The instrumentation and set points derived for this EAL set are consistent with the overall EAL scheme development guidance, address the plant-specific implementation strategies provided, and are consistent with a standard EAL scheme, as required by 1 O CFR 50.47(b)(4).

The NRC staff concludes that the plant-specific implementation method for this EAL set is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.2.2 EAL CA2/CU2 [CA2/CU2]

This EAL set is based upon a loss of available power to emergency power electrical busses. The NRC staff reviewed the licensee's evaluation and justification for plant-specific EAL changes associated with this set and has determined that the progression from UE to Alert is appropriate and consistent with EAL scheme development guidance.

The SAE and GE classification levels for this specific accident progression are bounded by indications available in EALs RS1 and RG1. Based on its review, the NRC staff has also determined that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance and with the plant-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The instrumentation and set points derived for this EAL set are consistent with the overall EAL scheme development guidance, address the plant-specific implementation strategies provided, and are consistent with a standard EAL scheme, as required by 10 CFR 50.47(b)(4). The NRG staff concludes that the plant-specific implementation method for this EAL set is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.2.3 EAL Set CA3/CU3 [CA3/CU3]

This EAL set is based upon an inability to maintain control of decay heat removal. The NRG staff reviewed the licensee's evaluation and justification for plant-specific EAL changes associated with this set and has determined that the progression from UE to Alert is appropriate and consistent with EAL scheme development guidance.

The SAE and GE classification levels for this specific accident progression are bounded by indications available in EALs RS1 and RG1. Based on its review, the NRG staff has also determined that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance and with the plant-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The instrumentation and set points derived for this EAL set are consistent with the overall EAL scheme development guidance, address the plant-specific implementation strategies provided, and are consistent with a standard EAL scheme, as required by 10 CFR 50.47(b)(4).

The NRG staff concludes that the plant-specific implementation method for this EAL set is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.2.4 EAL CU4 [CU4] This EAL is not part of an EAL set within the overall EAL scheme. The EAL's intent is to ensure that an EAL is declared when a loss of direct current (DC) power event occurs, as this condition compromises the ability of the licensee to monitor and control the removal of decay heat during cold shutdown or refueling modes of operation.

The Alert, SAE, and GE classification levels for this specific accident progression are bounded by indications available in EALs RA 1, RS1, and RG1. Based on its review, the NRG staff has determined that the numbering, sequencing, and format of this EAL are consistent with the overall EAL scheme development guidance and with the plant-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The instrumentation and set points derived for this EAL are consistent with the overall EAL scheme development guidance, address the plant-specific implementation strategies provided, and are consistent with a standard EAL scheme, as required by 1 O CFR 50.47(b)(4). The NRC staff concludes that the plant-specific implementation method for this EAL is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.2.5 EAL CU5 [CU5] This EAL is not part of an EAL set within the overall EAL scheme. The EAL's intent is to highlight the importance of emergency communications by ensuring that an EAL is declared if normal communication methods for onsite and offsite personnel or for offsite response organizations, including the NRC, are lost. The NRC staff reviewed the licensee's evaluation and justification for plant-specific changes associated with this EAL and has determined that no escalation path is necessary for this EAL. Based on its review, the NRC staff has also determined that the numbering, sequencing, and format of this EAL are consistent with the overall EAL scheme development guidance and with the plant-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The communication methods derived for this EAL are consistent with the overall EAL scheme development guidance, address the plant-specific implementation strategies provided, and are consistent with a standard EAL scheme, as required by 1 O CFR 50.47(b)(4).

The NRC staff concludes that the plant-specific implementation method for this EAL is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.2.6 EAL CA6 [CA6] This EAL is not part of an EAL set within the overall EAL scheme. The EAL's intent is to ensure that an EAL is declared when hazardous events lead to potential damage to safety systems. The SAE and GE classification levels for this accident progression are bounded by indications available in EALs RS1 and RG1. Based on its review of the proposed EAL, the NRC staff has determined that the numbering, sequencing, and format of this EAL are consistent with the overall EAL scheme development guidance and with the plant-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The NRC staff concludes that the plant-specific implementation method for this EAL is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b4), and is, therefore, acceptable for implementation. 3.3 Category 'E' -Independent Spent Fuel Storage Installation (ISFSI) 3.3.1 EALEU1 [E-HU1] This EAL is not part of an EAL set within the overall EAL scheme. The EAL's intent is limited to radiological events at the ISFSI. The NRC staff reviewed the licensee's evaluation and justification for plant-specific changes associated with this EAL and has determined that while security-related events at the ISFSI are also of concern, they are bounded by the licensee's EAL HA1. Based on its review, the NRC staff has also determined that the numbering, sequencing, and format of this EAL are consistent with the overall EAL scheme development guidance and with the plant-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The NRC staff concludes that the plant-specific implementation method for this EAL is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.4 Category 'F' -Fission Product Barrier Matrix This category is unique in the overall EAL scheme, as the thresholds are not intended to be stand-alone indicators of a particular event occurring at the plant. Rather, they are to be used as triggers within the particular logic configuration needed to reflect a loss or potential loss of a fission product barrier. The U.S. nuclear power plants have three fission product barriers:

fuel cladding, the RCS, and the primary containment.

Licensees are to develop thresholds that provide EAL decision-makers input into making an event declaration based upon degradation of one or more of these fission product barriers.

There are numerous triggers used as logic inputs to decide on the appropriate classification based upon the number of loss and/or potential loss indicators that are triggered for each barrier. By design, these indicators are redundant with other similar indicators in the Category 'R' and Category 'M' EAL sets, due to the importance for licensees to be able to recognize reactor and/or fission product barrier events as timely as possible, using the best available indicators from several different perspectives.

The NRC staff verified that the logic used to determine the appropriate emergency classification is consistent with the generic EAL scheme development guidance.

The NRC staff also verified that the instrumentation and set points derived for this EAL category are consistent with the overall EAL scheme development guidance, address the plant-specific implementation strategies provided, and are consistent with a standard EAL scheme, as required by 10 CFR 50.47(b)(4).

The licensee chose to modify this EAL category by using a plant-specific implementation method that uses a modified numbering format and EAL sequence other than that provided in the generic EAL scheme development guidance.

However, based on its review, the NRC staff has determined that the numbering, sequencing, and format of this EAL category are consistent with the overall EAL scheme development guidance and with the plant-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The NRG staff concludes that the plant-specific implementation method for this EAL category is in alignment with the key characteristics of an effective EAL scheme and, while different than that provided in the generic EAL development guidance, continues to meet the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.5 Category 'H' -Hazards 3.5.1 EAL Set HG1/HS1/HA1/HU1

[HG1/HS1/HA1/HU1]

This EAL set is based upon security-related events originally developed in accordance with the guidance from NRG Bulletin 2005-02 and RIS 2006-12 for licensees to implement, regardless of the specific version of the generic EAL scheme development guidance used, or if the particular licensee developed its EAL scheme using an alternative approach.

Based upon lessons learned from the implementation and use of this EAL set, particularly the insights gained from combined security and emergency preparedness drills, the NRG staff and the industry worked to enhance the language of these EALs so as to eliminate any confusion without changing the intent of the EAL set as set forth in NRG Bulletin 2005-02 and RIS 2006-12. Based on its review, the NRG staff has determined that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance and with the plant-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The NRG staff has also determined that this EAL set is consistent with the guidance provided in NRG Bulletin 2005-02 and RIS 2006-12, as further enhanced by the lessons learned from implementation and drills, and revised in NEI 99-01, Revision 6. The NRG staff concludes that the plant-specific implementation method for this EAL set is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.5.2 EAL HU2 [HU2] This EAL is not part of an EAL set within the overall EAL scheme. This EAL is based upon the effect that a seismic event may have on the facility.

The Alert, SAE, and GE classification levels for this specific accident progression are bounded by indications available in the fission product barrier matrix, as well as in EALs RA 1, RS 1, RG 1, CA6, and SAS. Based on its review, the NRG staff has determined that the numbering, sequencing, and format of this EAL are consistent with the overall EAL scheme development guidance and with the plant-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The NRC staff concludes that the plant-specific implementation method for this EAL is in alignment with the key characteristics of an effective EAL scheme, and while different than that provided in the generic EAL development guidance, it continues to meet the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.5.3 EAL HU3 [HU3] This EAL is not part of an EAL set within the overall EAL scheme. This EAL is based upon the effect that natural and destructive hazards may have on the facility.

The Alert, SAE, and GE classification levels for this specific accident progression are bounded by indications available in the fission product barrier matrix, as well as in EALs RA 1, RS 1, RG 1, CA6, and SAS. Based on its review, the NRC staff has determined that the numbering, sequencing, and format of this EAL are consistent with the overall EAL scheme development guidance and with the plant-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The instrumentation and set points derived for this EAL are consistent with the overall EAL scheme development guidance, address the plant-specific implementation strategies provided, and are consistent with a standard EAL scheme, as required by 1 O CFR 50.47(b)(4).

The NRC staff concludes that the plant-specific implementation method for this EAL is in alignment with the key characteristics of an effective EAL scheme and, while different than that provided in the generic EAL development guidance, continues to meet the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.5.4 EAL HU4 [HU4] This EAL is not part of an EAL set within the overall EAL scheme. This EAL is based upon the effect that fires may have on the facility.

The Alert, SAE, and GE classification levels for this specific accident progression are bounded by indications available in the fission product barrier matrix, as well as in EA Ls RA 1, RS 1, RG 1, CA6, and SA8. Based on its review, the NRC staff has determined that the numbering, sequencing, and format of this EAL are consistent with the overall EAL scheme development guidance and with the plant-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The NRC staff concludes that the plant-specific implementation method for this EAL is in alignment with the key characteristics of an effective EAL scheme and, while different than that provided in the generic EAL development guidance, continues to meet the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation. 3.5.5 EAL HA5 [HA5] This EAL is not part of an EAL set within the overall EAL scheme. This EAL is based upon the effect that toxic, corrosive, asphyxiant, or flammable gases may have on the facility.

The SAE and GE classification levels for this specific accident progression are bounded by indications available in the fission product barrier matrix, as well as in EALs RS1 and RG1. Based on its review, the NRG staff has determined that the numbering, sequencing, and format of this EAL are consistent with the overall EAL scheme development guidance and with the plant-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The NRG staff concludes that the plant-specific implementation method for this EAL is in alignment with the key characteristics of an effective EAL scheme and, while different than that provided in the generic EAL development guidance, continues to meet the requirements of Section IV of Appendix E to 10 GFR Part 50 and 10 GFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.5.6 EAL Set HS6/HA6 [HS6/HA6]

This EAL set is based upon control room evacuation and the inability to control critical plant systems remotely.

The NRG staff reviewed the licensee's evaluation and justification for specific changes associated with this EAL set and has determined that the progression from Alert to SAE is appropriate and consistent with EAL scheme development guidance.

The GE classification level for this specific accident progression is bounded by indications available in the fission product barrier matrix, as well as in EAL RG1. Based on its review, the NRG staff has also determined that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance and with the plant-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The NRG staff concludes that the plant-specific implementation method for this EAL set is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 1 O GFR Part 50 and 1 O GFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.5.7 EAL Set HG7/HS7/HA7/HU7

[HG7/HS7/HA7/HU7]

This EAL set is based upon providing the decision-makers with EALs to consider when, in their judgment, an emergency classification is warranted.

The NRG staff reviewed the licensee's evaluation and justification for plant-specific changes associated with this EAL set and has determined that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance, are consistent with the plant-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The NRC staff concludes that the plant-specific implementation method for this EAL set is in alignment with the key characteristics of an effective EAL scheme and, while different than that provided in the generic EAL development guidance, continues to meet the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.6 Category 'S' -System Malfunction 3.6.1 EAL Set SG1/SS1/SA1/SU1

[SG1/SS1/SA1/SU1]

This EAL set is based upon a loss of available alternating current (AC) power sources to the emergency busses. The NRC staff reviewed the licensee's evaluation and justification for plant-specific changes associated with this EAL set and has determined that the progression from UE to GE is appropriate and consistent with EAL scheme development guidance.

The licensee chose to modify this EAL set by using a site-specific implementation method that uses a modified numbering format and EAL sequence other than that provided in the generic EAL scheme development guidance.

However, based on its review, the NRC staff has determined that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance and with the site-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The instrumentation, values, and listing of applicable power sources derived for this EAL set are consistent with the overall EAL scheme development guidance, address the plant-specific implementation strategies provided, and are consistent with a standard EAL scheme, as required by 10 CFR 50.47(b)(4).

The NRC staff concludes that the plant-specific implementation method for this EAL set is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.6.2 EAL Set SG1 .2/SS2 [SG8/SS8]

This EAL set is based upon a loss of site AC and DC power sources. The EAL's intent is to ensure that an EAL is declared when a loss of AC or DC power event occurs, as this condition compromises the ability of the licensee to monitor and control the removal of decay heat. The licensee chose to modify this EAL set by using a site-specific implementation method that uses a modified numbering format and EAL sequence other than that provided in the generic EAL scheme development guidance.

However, based on its review, the NRC staff has determined that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance and with the site-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The NRC staff concludes that the plant-specific implementation method for this EAL is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.6.3 EAL Set SA3/SU3 [SA2/SU2]

This EAL set is based upon the effect that a loss of available indicators in the control room has on the facility.

The NRC staff has reviewed the licensee's evaluation and justification for specific changes associated with this EAL set and determined that the progression from UE to Alert is appropriate and consistent with EAL scheme development guidance.

The SAE and GE classification levels for this specific accident progression are bounded by indications available in the fission product barrier matrix, as well as in EALs RS1 and RG1. The licensee chose to modify this EAL set by using a site-specific implementation method that uses a modified numbering format and EAL sequence other than that provided in the generic EAL scheme development guidance.

However, based on its review, the NRC staff has determined that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance and with the site-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The instrumentation and set points derived for this EAL set are consistent with the overall EAL scheme development guidance, address the plant-specific implementation strategies provided, and are consistent with a standard EAL scheme, as required by 1 O CFR 50.47(b)(4).

The NRC staff concludes that the plant-specific implementation method for this EAL set is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.6.4 EAL SUS [SU4] This EAL is not part of an EAL set within the overall EAL scheme. The EAL's intent is to ensure that an EAL is declared when the plant has indications of RCS leakage. By design, this EAL is redundant with corresponding indicators from a loss or potential loss of fission product barriers, as well as radiation monitoring, to ensure reactor and/or fission product barrier events are recognized regardless of the particular EAL table a licensee may be referring to. EAL escalation is bounded by indications available in the fission product barrier matrix, as well as in EALs RA1, RS1, and RG1. The licensee chose to modify this EAL set by using a site-specific implementation method that uses a modified numbering format and EAL sequence other than that provided in the generic EAL scheme development guidance.

However, based on its review, the NRC staff has determined that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance and with the site-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The NRC staff concludes that the plant-specific implementation method for this EAL is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.6.5 EAL Set SS6/SA6/SU6

[SS5/SA5/SU5]

This EAL set is based upon the effect that a failure of the reactor protection system may have on the plant. The NRC staff reviewed the licensee's evaluation and justification for specific changes associated with this EAL set and has determined that the progression from UE to SAE is appropriate and consistent with EAL scheme development guidance.

The GE classification level for this event is bounded by indications available in the fission product barrier matrix, as well as in EAL RG 1 . The licensee chose to modify this EAL set by using a site-specific implementation method that uses a modified numbering format and EAL sequence other than that provided in the generic EAL scheme development guidance.

However, based on its review, the NRC staff has determined that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance and with the site-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The instrumentation and set points derived for this EAL set are consistent with the overall EAL scheme development guidance, address the plant-specific implementation strategies provided, and are consistent with a standard EAL scheme, as required by 10 CFR 50.47(b)(4).

The NRC staff concludes that the plant-specific implementation method for this EAL set is in alignment with the key characteristics of an effective EAL scheme and, while different than that provided in the generic EAL development guidance, continues to meet the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.6.6 EAL SU7 [SU6] This EAL is not part of an EAL set within the overall EAL scheme. The EAL's intent is to highlight the importance of emergency communications by ensuring that an EAL is declared if normal communication methods for onsite and offsite personnel or for offsite response organizations, including the NRC, are lost. The NRC staff reviewed the licensee's evaluation and justification for plant-specific changes associated with this EAL and has determined that no escalation path is necessary.

The licensee chose to modify this EAL set by using a site-specific implementation method that uses a modified numbering format and EAL sequence other than that provided in the generic EAL scheme development guidance.

However, based on its review, the NRC staff has determined that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance and with the site-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The communication methods derived for this EAL are consistent with the overall EAL scheme development guidance, address the plant-specific implementation strategies provided, and are consistent with a standard EAL scheme, as required by 10 CFR 50.47(b)(4). The NRC staff concludes that the plant-specific implementation method for this EAL is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.6.7 EAL SU4 [SU3] This EAL is not part of an EAL set within the overall EAL scheme. The EAL's intent is to ensure that an EAL is declared when RCS activity is greater than technical specification allowable limits. The Alert, SAE, and GE classification levels for this specific accident progression are bounded by indications available in the fission product barrier matrix, as well as in EALs RA 1, RS1, and RG1. The licensee chose to modify this EAL set by using a site-specific implementation method that uses a modified numbering format and EAL sequence other than that provided in the generic EAL scheme development guidance.

However, based on its review, the NRC staff has determined that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance and with the site-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The NRC staff concludes that the plant-specific implementation method for this EAL is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation.

3.6.8 EAL SAS [SA9] This EAL is not part of an EAL set within the overall EAL. The EAL's intent is to ensure that an EAL is declared when hazardous events lead to potential damage to safety systems. The SAE and GE classification levels for this accident progression are bounded by indications available in the fission product barrier matrix, as well as in EALs RS1 and RG1. The licensee chose to modify this EAL set by using a site-specific implementation method that uses a modified numbering format and EAL sequence other than that provided in the generic EAL scheme development guidance.

However, based on its review, the NRC staff has determined that the numbering, sequencing, and format of this EAL set are consistent with the overall EAL scheme development guidance and with the site-specific implementation strategies provided, and are, therefore, considered part of a standard EAL scheme. The NRC staff concludes that the plant-specific implementation method for this EAL is in alignment with the key characteristics of an effective EAL scheme, meets the requirements of Section IV of Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(4), and is, therefore, acceptable for implementation. 3.7 Review Result Summary The NRC staff has reviewed the technical bases for the proposed EAL scheme, the modifications from NEI 99-01, Revision 6, and the licensee's evaluation of the proposed changes. The licensee chose to modify its proposed EAL scheme from the generic EAL scheme development guidance provided in NEI 99-01, Revision 6, in order to adopt a format that is better aligned with how it currently implements its EALs, as well as with plant-specific writer's guides and preferences.

The NRC staff determined that these modifications do not alter the intent of any specific EAL within a set, category, or within the entire EAL scheme described in NEI 99-01, Revision 6. Thus, the proposed changes meet the requirements in Appendix E to 10 CFR Part 50 and the planning standards of 10 CFR 50.47(b).

Therefore, the NRC staff concludes that the licensee's proposed EAL scheme is acceptable and provides reasonable assurance that the licensee can and will take adequate protective measures in the event of a radiological emergency.

Specifically, the staff concludes that the licensee's site-specific EAL basis document provided by Enclosure 3 of the letter dated November 23, 2015, is acceptable for implementation.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the North Carolina official was notified on December 24, 2015, of the proposed issuance of the amendment.

The state official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION The amendment changes a requirement with respect to the installation or use of facility components located within the restricted area as defined in 1 O CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure.

The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on April 28, 2015 (80 FR 23602). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 1 O CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1. Generic Letter 79-50 dated October 10, 1979 (ADAMS Accession No. ML031320278).

2. U.S. Nuclear Regulatory Commission and Federal Emergency Management Agency, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," NUREG-0654/FEMA-REP-1, Revision 1, November 1980 (ADAMS Accession No. ML040420012).

3. Nuclear Management and Resources Council/National Environmental Studies Project (NUMARC/NESP)-007, Revision 2, "Methodology for Development of Emergency Action Levels," January 1992 (ADAMS Accession No. ML041120174).

4. NEI 99-01, Revision 4, "Methodology for Development of Emergency Action Levels," January 2003 (ADAMS Accession No. ML041470143).

5. NEI 99-01, Revision 5, "Methodology for Development of Emergency Action Levels," February 2008 (ADAMS Accession No. ML080450149).

6. NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ADAMS Package Accession No. ML13091A209). 7. Thaggard, M., U.S. Nuclear Regulatory Commission, Letter to Ms. Perkins-Grew, Nuclear Energy Institute, "U.S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 6, Dated November, 2012," March 28, 2013 (ADAMS Accession No. ML12346A463).

8. U.S. Nuclear Regulatory Commission, "Emergency Planning and Preparedness for Nuclear Power Reactors," Regulatory Guide 1.101, Revision 2, October 1981 (ADAMS Accession No. ML090440294);

Revision 3, August 1992 (ADAMS Accession No. ML003740302);

and Revision 4, July 2003 (ADAMS Accession No. ML032020276).

9. Miller, C. G., U.S. Nuclear Regulatory Commission, Letter to Alan Nelson, Nuclear Energy Institute, "U.S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 5, Dated February 2008," February 22, 2008 (ADAMS Accession No. ML080430535).

10. U.S. Nuclear Regulatory Commission, Regulatory Issue Summary 2003-18, with Supplements 1 and 2, "Use of NEI 99-01, 'Methodology for Development of Emergency Action Levels,' Revision 4, Dated January 2003, October 8, 2003 (ADAMS Accession Nos. ML032580518, ML041550395, and ML051450482).

11. NRC Bulletin 2005-02, "Emergency Preparedness and Response Actions for Security-Based Events,

July 18, 2005 (ADAMS Accession No. ML051740058).

12. NRC Regulatory Issue Summary 2006-12, "Endorsement of Nuclear Energy Institute Guidance 'Enhancements to Emergency Preparedness Programs for Hostile Action,"'

July 19, 2006 (ADAMS Accession No. ML072670421

). 13. Letter from Duke Energy Progress, Inc., to U.S. Nuclear Regulatory Commission, "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',

January 30, 2015 (ADAMS Accession No. ML15044A198). 14. Letter from Duke Energy Progress, Inc., to U.S. Nuclear Regulatory Commission, "License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, 'Development of Emergency Action Levels for Non-Passive Reactors'," November 23, 2015 (ADAMS Package Accession No. ML15350A105).

15. NRC Regulatory Issue Summary 2003-18, with Supplements 1 and 2, "Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels" (ADAMS Accession No. ML051180469).

Principal Contributor:

Don Johnson, NSIR Dated: January 8, 2016 W. Gideon A Notice of Issuance will be included in the Commission's Biweekly Federal Register Notice. Sincerely, IRA/ Andrew Hon, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-324 and 50-325

Enclosures:

1. Amendment No. 268 to DPR-71 2. Amendment No. 296 to DPR-62 3. Safety Evaluation DISTRIBUTION:

PUBLIC: (No PDC/Listserv for 10 working days) LPL2-2 R/F RecordsAmend RidsNrrLABClayton Resource RidsACRS_MailCTR Resource RidsNsirDprOrlob Resource RidsNrrDorlDpr Resource RidsNrrPMBrunswick Resource RidsNrrDorllpl2-2 Resource DJohnson, NSIR RidsNrrDssSrxb Resource RidsRgn2MailCenter Resource CPfefferkorn, NRR ADAMS Accession No.: ML15344A153

• by memo (ML15303A396)

OFFICE NRR/DORL/LPLI 1-2/PI NRR/DORL/LPLll-2/LA NSIR/DPR/ORLOB/BC*

NAME AH on (LRonewicz for) BClayton JAnderson DATE 12/16/2016 12/16/2015 12/08/2015 OFFICE OGC-NLO N RR/DORL/LPLI 1-2/BC NRR/DORL/LPLll-2/PM NAME BMizuno BBeasley AH on DATE 1/05/2016 1/08/2016 1/08/2016 OFFICIAL RECORD COPY