View source for NEI 06-02, License Amendment Request (LAR) Guidelines

[[NEI::NEI 06-02|NEI 06-02]] [Revision 5]
LICENSE AMENDMENT REQUEST ([[LAR|LAR]]) GUIDELINES
December 2016
Nuclear Energy Institute, 1201 F St., NW, Suite 1100, Washington, [[DC|DC]] 20004-1218(202.739.8000)
[[NEI::NEI 06-02|NEI 06-02]] [Revision 5]
Nuclear Energy Institute
LICENSE AMENDMENT REQUEST ([[LAR|LAR]]) GUIDELINES
December 2016
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ACKNOWLEDGEMENTS
The NEI Regulatory Issues Task Force wishes to acknowledge the work of the Pressurized Water Reactor Owners Group ([[PWROG|PWROG]]), the Boiling Water Reactor Owners' Group ([[BWROG|BWROG]]), the AP1000® Owners Group ([[APOG|APOG]]), and the NEI Regulatory Issues Task Force (RITF) in revising [[NEI::NEI 06-02|NEI 06-02]] to incorporate improvements identified as part of the industry Delivering the Nuclear Promise initiative.
NOTICE
Neither NEI, nor any of its employees, members, supporting organizations, contractors, or consultants make any warranty, expressed or implied, or assume any legal responsibility for the accuracy or completeness of, or assume any liability for damages resulting from any use of, any information apparatus, methods, or process disclosed in this report or that such may not infringe privately owned rights.
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==ABSTRACT==
[[NEI::NEI 06-02|NEI 06-02]], Revision 5, describes a standardized approach to the license amendment process used by commercial nuclear power plant licensees.1
This document describes the license amendment process and provides a standard format for plant-specific license amendment requests.
The appendices provide detailed information on specific subjects, such as:
* Exigent/Emergency [[LAR|LAR]]s
* Plant-specific Adoption of [[TSTF|TSTF]] Travelers
* License Amendment Requests with Risk-Informed Justification
* Pre-Submittal Meeting Guidance
* Voluntary vs. Non-Voluntary License Amendment Requests
* Industry Coordinated and Multi-Licensee Consolidated License Amendment Requests
1Disclaimer – Discussions of NRC activities in this document are based on industry understanding of NRC activities. The NRC staff follows their internal guidance documents.
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TABLE OF CONTENTS
1 THE LICENSE AMENDMENT REQUEST PROCESS ...................................................... 1
2 INITIATION ................................................................................................................... 2
3 PRE-SUBMITTAL MEETINGS ........................................................................................ 4
4 PREPARATION ............................................................................................................. 5
5 VERIFICATION ............................................................................................................. 8
6 IDENTIFYING CHANGE IMPACTS ................................................................................ 9
7 PEER REVIEW .............................................................................................................. 9
8 TECHNICAL REVIEW .................................................................................................. 10
9 CONCURRENCE REVIEW .......................................................................................... 11
10 SAFETY COMMITTEE REVIEW ................................................................................... 12
11 APPROVAL AND SUBMITTAL .................................................................................... 12
12 NRC ACCEPTANCE REVIEW ...................................................................................... 14
13 NRC REQUESTS [[FOR|FOR]] ADDITIONAL INFORMATION PROCESS AND AUDITS ........... 15
14 NRC APPROVAL, REJECTION, [[OR|OR]] WITHDRAWAL .................................................... 16
15 IMPLEMENTATION ..................................................................................................... 17
APPENDIX A STANDARD FORMAT [[FOR|FOR]] PLANT-SPECIFIC LICENSE AMENDMENT REQUESTS ................................................................................................................ A-1
APPENDIX B [[LAR|LAR]] EXAMPLE ............................................................................................ B-1
APPENDIX C EXIGENT/EMERGENCY LARS .................................................................... C-1
APPENDIX D PLANT-SPECIFIC ADOPTION OF [[TSTF|TSTF]] TRAVELERS ................................. D-1
APPENDIX E LICENSE AMENDMENT REQUESTS WITH RISK-INFORMED JUSTIFICATION ......................................................................................................... E-1
APPENDIX F PRE-SUBMITTAL MEETING GUIDANCE ...................................................... F-1
ATTACHMENT – PRE-SUBMITTAL GUIDANCE CHECKLIST...............................................G-5
APPENDIX G VOLUNTARY VS. NON-VOLUNTARY LICENSE AMENDMENT REQUESTS G-1
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APPENDIX H INDUSTRY CONSOLIDATED AND MULTI-LICENSEE COORDINATED LICENSE AMENDMENT REQUESTS ......................................................................... H-1
APPENDIX I GLOSSARY .................................................................................................... I-1
APPENDIX J [[Acronym|ACRONYM]]S ................................................................................................. J-1
APPENDIX K REFERENCES ............................................................................................. K-1
[[NEI::NEI 06-02|NEI 06-02]], Revision 5
December 2016
1
LICENSE AMENDMENT REQUEST ([[LAR|LAR]]) GUIDELINES

==1 THE LICENSE AMENDMENT REQUEST PROCESS==
The following sections discuss the recommended industry process for [[LAR|LAR]]s, which was developed to assure quality, efficiency, and consistency. Licensees are encouraged to use the nomenclature and format of the recommended [[LAR|LAR]] process and example documents to increase consistency across the industry, to allow the efficient use of generic products, and to facilitate the use of industry resources to develop [[LAR|LAR]]s. Each section discusses the overall objective of that step, background information describing the step, and specific activities that should be performed by the licensee during that step of the [[LAR|LAR]] process. Provided that the objective of a step is met, licensees may alter the order of performance of steps, or portions of the steps, and develop forms, checklists, software, etc. to fit their organization's needs. The background description of each step is not required to be included in the [[LAR|LAR]] process.
This guideline considers three types of [[LAR|LAR]]s:
Type 1. A plant-specific change for which there is no Nuclear Regulatory Commission (NRC) accepted model application.
An example of a Type 1 [[LAR|LAR]] is a plant-specific setpoint change.
Type 2. A change based on an NRC-approved generic justification, such as a Technical Specifications Task Force ([[TSTF|TSTF]]) traveler, for which there is an NRC-accepted model application that requires submittal of technical, plant-specific information.
An example of a Type 2 [[LAR|LAR]] is adoption of [[TSTF::TSTF-425|TSTF-425]], "Relocate Surveillance Frequencies to Licensee Control - [[RITSTF|RITSTF]] Initiative 5b." The NRC-approved model application for [[TSTF::TSTF-425|TSTF-425]] requires inclusion of a description of the probabilistic risk assessment ([[PRA|PRA]]) model adequacy to support the change.
Type 3. A change based on an NRC approved generic justification, such as a [[TSTF|TSTF]] traveler, for which there is an NRC-accepted model application that does not require submittal of technical, plant-specific information other than choosing the appropriate plant specific options provided in the model application.
An example of a Type 3 [[LAR|LAR]] is adoption of [[TSTF::TSTF-545|TSTF-545]], "TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing." The NRC-approved model application for [[TSTF::TSTF-545|TSTF-545]] does not require the inclusion of any plant-specific technical information.
Note that plant-specific system names or Technical Specification (TS) numbers are not considered technical information unless they affect the applicability of the generic justification.

Each [[LAR|LAR]] type is subjected to utility internal reviews prior to submittal to the NRC for review and approval. Utility internal [[LAR|LAR]] review typically consists of four distinct activities:
* Peer review by RA2 to verify that all regulatory requirements are met.
* Technical review by the responsible organizations to verify that all technical information is accurate and complete.
* Management review to verify the change is consistent with company goals and objectives.
* Safety committee review to independently verify the proposed change does not have a detrimental effect on nuclear safety.
It is important that the above review activities remain separate to avoid the unintentional omission of a needed review; such as a management reviewer performing a technical review instead of considering whether the change is consistent with company goals and objectives.

==2 INITIATION==
Objective
To define a process for requesting a [[LAR|LAR]] or identifying that a [[LAR|LAR]] is required, prioritizing the [[LAR|LAR]] requests, obtaining management agreement to pursue the change, and tracking the progress of planned [[LAR|LAR]]s. Note: Prioritization and sequencing of [[LAR|LAR]] submittals is determined by the utility.
Background
Commercial nuclear power plant activities that involve certain types of changes, such as a change to the Technical Specifications (TS), system modifications, or changes to operating procedures, may require NRC approval. The following regulations are examples of those that describe when prior NRC approval in accordance with [[CFR::10 CFR 50.90#|10 CFR 50.90]] is necessary:
* [[CFR::10 CFR 50.90#|10 CFR 50.90]] specifies the change process for the Operating License ([[OL|OL]]) and TS.
* [[CFR::10 CFR 50.54#p|10 CFR 50.54(p)]] specifies the change process for the Safeguards Contingency Plan, Security Plan, and the guard training and qualification plan.
* [[CFR::10 CFR 50.54#a|10 CFR 50.54(a)]] specifies the change process for the Quality Assurance plan.
2 This document uses the [[Acronym|acronym]] "[[RA|RA]]" to signify the organizational name "Regulatory Affairs" or "Regulatory Assurance" as the organization with primary responsibility for managing the [[LAR|LAR]] process. The title "Licensing" is also often used. The organization name is not part of the recommended [[LAR|LAR]] process.

* [[CFR::10 CFR 50.54#q|10 CFR 50.54(q)]] specifies the change process for the Emergency Plan.
* [[CFR::10 CFR 50.59#|10 CFR 50.59]] specifies the change process for changes, tests, and experiments.
Change control processes are discussed in several licensee programs, industry guidelines, and Nuclear Reactor Regulation ([[NRR|NRR]]) Office Instructions related to [[LAR|LAR]]s, such as:
* Technical Specification Bases Control Program, which is a licensee-controlled program described in the Administrative Controls chapter of the TS for managing the content of the Bases of the plant-specific TS.
* [[NEI::NEI 96-07|NEI 96-07]], Revision 1, "Guidelines for [[CFR::10 CFR 50.59#|10 CFR 50.59]] Implementation" (Ref. 1), which is endorsed by NRC in [[Regulatory Guide::Regulatory Guide 1.187|Regulatory Guide 1.187]] (Ref. 2).
* [[NEI::NEI 99-04|NEI 99-04]], "Guidelines for Managing NRC Commitment Changes" (Ref. 3), which is endorsed by the NRC in Regulatory Issue Summary ([[RIS|RIS]]) 2000-17 (Ref. 4).
* [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-100|LIC-100]] (Ref. 5), which is the NRC's internal guidance on the terminology and documents associated with the licensing bases for an operating nuclear power plant.
* [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-101|LIC-101]] (Ref. 6), which is the NRC's internal guidance on the license amendment review process.
* [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-109|LIC-109]] (Ref. 7), which is the NRC's internal guidance on the initial acceptance review process for [[LAR|LAR]]s.
If a proposed activity requires prior NRC approval of a change to the [[OL|OL]], the licensee must submit a [[LAR|LAR]] to the NRC in accordance with [[CFR::10 CFR 50.90#|10 CFR 50.90]]. Additional requirements pertaining to [[LAR|LAR]]s are contained in [[CFR::10 CFR 50.91#|10 CFR 50.91]] and [[CFR::10 CFR 50.92#|10 CFR 50.92]].
Process
a. Licensees use varying methods to identify the need for, and track submittal of [[LAR|LAR]]s. Some licensees have procedural guidance for requesting a change to any licensing basis document (TS, TS Bases, Final Safety Analysis Report ([[FSAR|FSAR]]), etc.) Other licensees use a change request process for each document or groups of documents, and some licensees do not have a formal process other than an action tracking database. There is no regulatory requirement for a licensee to track future [[LAR|LAR]] submittals (other than the requirement to resolve [[Conditions Adverse to Quality|conditions adverse to quality]], such as nonconservative TS) and, therefore, each licensee may manage how [[LAR|LAR]]s are identified for submittal in the manner they deem best supports their organizational goals.
b. It is recommended that licensees have a mechanism for tracking [[LAR|LAR]]s that are planned to be submitted. In Regulatory Issue Summary ([[RIS|RIS]]) 2015-16, Revision 1, "Planned Licensing Action Submittals for All Power Reactor Licensees," the NRC requested that licensees provide information regarding the licensing actions they plan to submit to the NRC for review over the next two calendar years, and power uprate applications they plan to submit to the NRC for review over the next five calendar years. All licensees responded and the industry agreed to maintain a current list of planned [[LAR|LAR]]s with the NRC Project Managers.
c. Assignment of a principal person responsible for the creation of a [[LAR|LAR]] is the responsibility of the [[RA|RA]] organization management. The [[RA|RA]] organization is not governed by the Institute of Nuclear Power Operations ([[INPO|INPO]]), [[ACAD|ACAD]] 02-002, "The Process for Maintaining Accreditation of Training in the Nuclear Power Industry Technical Areas." Therefore, formal qualification programs for [[LAR|LAR]] preparers, such as Qualification Cards or Job Performance Measures, are not recommended. However, it is recognized that training and mentoring are vital to an effective [[RA|RA]] organization and that the [[RA|RA]] organization management is responsible for assigning personnel with the requisite knowledge to prepare a particular [[LAR|LAR]]. [[RA|RA]] organization management may also choose to have third parties (consultants, Owners Groups, NEI, etc.) prepare a [[LAR|LAR]] based on the knowledge and experience of the responsible preparer.
d. Assignment of a management sponsor for some or all [[LAR|LAR]]s to ensure station resources are assigned to support the request is a utility-specific choice.

==3 PRE-SUBMITTAL MEETINGS==
Objective
To provide assurance that a submitted [[LAR|LAR]] will be acceptable for NRC review. Pre-submittal meetings are optional activities that are typically requested by the licensee for complex submittals to ensure agreement with the NRC staff on applicable regulatory requirements, technical content, necessary level of detail, use of precedent or topical reports, and schedule.
Background
Licensees may request a pre-submittal meeting (which may be a conference call in lieu of a physical meeting) to facilitate the submittal of a [[LAR|LAR]] acceptable for NRC review. Pre-submittal meetings can be especially helpful in complex, "first of a kind" ([[FOAK|FOAK]]), and digital upgrade [[LAR|LAR]]s. For very complex changes, such as digital upgrade [[LAR|LAR]]s, more than one pre-submittal meeting may be necessary.
Although pre-submittal meetings can be useful for determining reasonable and acceptable approaches to a planned [[LAR|LAR]], licensees should take care to not ask questions that seek a determination from the NRC on an appropriate course of action. The licensee should clearly understand the goal of the meeting, such as guidance on level of detail, justification required for use of a new method, applicability of a precedent, feasibility of a desired schedule, etc.
In pre-submittal meetings, it's important to ask questions of the NRC staff and to not assume that no NRC questions mean staff agreement. In all cases, NRC staff statements in a pre-submittal meeting should not be taken as staff endorsement of a [[LAR|LAR]]. Generally, no regulatory decisions are made at pre-submittal meetings.

Process
Detailed pre-submittal meeting guidance and a checklist are provided in Appendix F, "Pre-submittal Meeting Guidance."
a. Pre-submittal meetings are optional activities and it is not necessary to hold a presubmittal meeting for all planned [[LAR|LAR]]s. Pre-submittal meetings are recommended for [[FOAK|FOAK]] and complex changes. For very complex changes, such as digital upgrade [[LAR|LAR]]s, more than one pre-submittal meeting may be necessary.
b. The licensee should work with the NRC [[PM|PM]] to identify all NRC groups that should be represented and to ensure that all necessary staff attends. The licensee should work with the NRC [[PM|PM]] to identify all NRC groups that should be represented. Omitting an NRC branch from the pre-submittal meeting may lead to acceptance concerns not being identified prior to the [[LAR|LAR]] submittal and may risk non-acceptance of the [[LAR|LAR]]. The licensee should consider requesting that the Branch Chief of the lead NRC technical organization(s) for the submittal attend to ensure continuity if there is a subsequent change in reviewer(s).
c. The licensee should work with the NRC [[PM|PM]] to determine the timing for submission of materials to the NRC to be used at the pre-submittal meeting in order to allow for the information to be placed in the NRC's Agency-wide Documents Access and Management System ([[ADAMS|ADAMS]]). This allows the NRC staff and members of the public participating by teleconference to view the materials. Ideally, handout materials should be provided prior to the preparation of the meeting announcement so that the NRC [[PM|PM]] and technical staff have adequate time to prepare for a productive meeting. The NRC has indicated that it is helpful to identify the plant-specific licensing basis requirements related to the submittal. If proprietary material needs to be discussed during the pre-submittal meeting, ensure that this is communicated to the NRC [[PM|PM]] so the meeting can be arranged to have a both a public portion and a closed portion.
d. At the end of the meeting, have a thorough closing in which all parties concur with the results of the meeting. Clearly document the results of the meeting, including any outstanding issues. Work with the [[PM|PM]] to issue a meeting summary. In the subsequent submittal, refer to the pre-submittal meeting and any documented meeting summary.

==4 PREPARATION==
Objective
To create a [[LAR|LAR]] and supporting documentation that is technically correct, complete, and encompasses all necessary regulatory requirements.
Background
Writing an effective [[LAR|LAR]] is a creative and iterative process. It requires drafting a narrative that effectively describes the current condition, the need for change, the proposed change, and the technical justification for the change, while meeting the regulatory and administrative requirements.
Process
a. For a Type 1 [[LAR|LAR]], the submittal should follow the standard format described in Appendix A. Appendix A provides an explanation for each element of the [[LAR|LAR]]. Appendix B provides an example [[LAR|LAR]] developed in accordance with the guidance of Appendix A. Other appendices discuss specific aspects of [[LAR|LAR]]s:
* Appendix C contains guidance for modifying the standard format to accommodate exigent or emergency circumstances.
* Appendix D contains guidance on [[LAR|LAR]]s based on [[TSTF|TSTF]] travelers with and without model applications.
* Appendix E contains guidance on [[LAR|LAR]]s with risk-informed justification.
* Appendix G contains guidance on non-voluntary [[LAR|LAR]]s.
b. For a Type 2 or Type 3 [[LAR|LAR]], the submittal should follow the NRC-approved model application as closely as the plant design and licensing basis allows. Changes not described in the "Variations" section of the model application should not be included. Changes not described in the NRC-approved model Safety Evaluation should not be included. Simple administrative (i.e., typographical corrections) or conforming changes needed to address differences between the plant-specific TS and the traveler markups are acceptable variations from the traveler. If in doubt, discuss the changes with the NRC [[PM|PM]]. If technically significant deviations from the generically approved change are needed, a Type 1 [[LAR|LAR]] should be written that does not claim to adopt the traveler (although the traveler may be cited as precedent for the applicable portions of the [[LAR|LAR]]). Appendix D contains guidance on [[LAR|LAR]]s based on [[TSTF|TSTF]] travelers with and without model applications.
c. In preparing a [[LAR|LAR]], identify and review relevant documents, such as:
* For a Type 2 or 3 change, the [[TSTF|TSTF]] traveler, any supporting topical report, and the NRC safety evaluation.
* Design basis, technical, licensing basis, or vendor documents supporting the change.
* Standard Technical Specifications ([[NUREG::NUREG-1430|NUREG-1430]] through -1434, and -2194). Identify differences between the affected plant-specific TS and the standard TS.
* Relevant approved, rejected, and draft [[TSTF|TSTF]] travelers for applicability to the proposed change.
* Relevant NRC regulatory guidance (Regulatory Guides, Standard Review Plan, Office Instructions, etc.)
* Relevant industry standards (NEI, IEEE, [[EPRI|EPRI]], [[ASME|ASME]], ANSI, etc.)
* Precedent [[LAR|LAR]]s from other licensees.
d. Document the review and the source of any information to be used in the [[LAR|LAR]] for future verification of [[LAR|LAR]] content.
e. Identify internal stakeholders and resources needed to develop and implement the [[LAR|LAR]], such as, Operations, Engineering, Outage Planning, and [[PRA|PRA]]. A stakeholder meeting is recommended to gain alignment on responsibilities and schedule.
f. If the [[LAR|LAR]] is an emergency or exigent [[LAR|LAR]], see Appendix C.
g. Determine the implementation period and document the basis of any extended period in the [[LAR|LAR]], recognizing that a change to the implementation date after approval of the [[LAR|LAR]] requires an additional [[LAR|LAR]].
h. If proprietary or sensitive information is included in the [[LAR|LAR]], refer to [[CFR::10 CFR 2.390#|10 CFR 2.390,]] [[Regulatory Issue Summary::RIS 2014-01|RIS 2014-01]], "Regulatory Requirements for Withholding of Proprietary Information from Public Disclosure," [[Regulatory Issue Summary::RIS 2004-11|RIS 2004-11]], "Supporting Information Associated with Requests for Withholding Proprietary Information," and [[NRC information notice::Information Notice 2009-07|IN 2009-07]], "Withholding of Proprietary Information from Public Disclosure," to ensure that proprietary information is withheld and the appropriate justification for withholding the information is provided in the affidavit and the correspondence. An affidavit must accompany any piece of regulatory correspondence in which it is requested to withhold certain information due to its proprietary nature or other confidential or privileged content per [[CFR::10 CFR 2.390#b1|10 CFR 2.390(b)(1)]]. [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-204|LIC-204]], "Handling Requests to Withhold Proprietary Information from Public Disclosure," Section 4.0 may be used as a guide for affidavit content and structure. For each document marked as containing proprietary information, provide a corresponding non-proprietary version of the document unless the document is considered proprietary in its entirety.
i. Documents submitted to the NRC that contain Sensitive Unclassified Nonsafeguards Information ([[SUNSI|SUNSI]]) should be prepared consistent with [[Regulatory Issue Summary::RIS 2005-26|RIS 2005-26]], "Control of Sensitive Unclassified Nonsafeguards Information Related to Nuclear Power Reactors." Information to be withheld from public disclosure in accordance with [[CFR::10 CFR 2.390#d1|10 CFR 2.390(d)(1)]] or other provision in the regulation should be subject to the same general identification practices as used for proprietary commercial or financial information. The cover letter should clearly state that the document includes sensitive information and the affected pages should include the marking "Security-Related Information Withhold Under [[CFR::10 CFR 2.390#|10 CFR 2.390]]." Unlike the requirements for withholding proprietary information, no affidavit is required for sensitive information withheld under [[CFR::10 CFR 2.390#d|10 CFR 2.390(d)]] and related to physical protection or material control and accounting.

==5 VERIFICATION==
Objective
To ensure the [[LAR|LAR]] is accurate in all material respects by confirming the [[LAR|LAR]] information against supporting documents. The verification package is used to support internal review, response to NRC questions, and to provide historical evidence.
Background
A [[LAR|LAR]] is "information provided to the Commission," and, as such, must be complete and accurate in all material respects, in accordance with [[CFR::10 CFR 50.9#|10 CFR 50.9]]. The statements of consideration for [[CFR::10 CFR 50.9#|10 CFR 50.9]] stated:
The Commission decided materiality is to be judged by whether information has a natural tendency or capability to influence an agency decisionmaker; that knowledge of the falsity of a material statement is not necessary for a material false statement under section 186 and that material omissions are actionable to the same extent as affirmative material false statements.
Both the utility and individual utility employees can be held legally accountable for "material false statements." A failure to provide complete and accurate information could be considered a material false statement if the inaccurate statement is made knowing the statement is inaccurate or incomplete, or with careless disregard for its accuracy or completeness; or it is made with a clearly demonstrable knowledge of its inaccuracy or incompleteness.
Process
a. In parallel with preparation of the [[LAR|LAR]], the preparer should follow the utility-specific process to identify the "verifiable statement(s)" in the [[LAR|LAR]]. Verifying the completeness and accuracy of the submitted information applies to [[LAR|LAR]] Types 1, 2, and 3.
b. A "verifiable statement" is information in the [[LAR|LAR]] that can be used by the NRC to make a regulatory decision. The extent of verifiable statements is a utility management decision. At a minimum, all information that is material to the review should be verified. Some utilities choose to consider all statements of fact in the [[LAR|LAR]] to be verifiable statements.
c. Provide objective evidence (e.g., design and licensing basis documents) that support the accuracy of each verifiable statement. Engineering judgement may be acceptable as objective evidence if documented according to utility-specific processes.
d. The method of documenting the verifiable statements is utility-specific. Examples are using the comments feature in Microsoft Word or using a highlighter and margin notes to annotate statements with supporting document references. It is recommended that the electronic or physical copy of the [[LAR|LAR]] with the identified verifiable statements and the referenced supporting documents be retained by the TA organization for the life of the plant to facilitate future research into the basis of the [[LAR|LAR]].

==6 IDENTIFYING CHANGE IMPACTS==
Objective
To identify and track the procedures, programs, testing, design changes, design documents, training, etc. that require revision or performance to support the approved amendment to ensure compliance with the approved change and performance of any required actions.
Background
Implementation of an approved [[LAR|LAR]] goes beyond inserting the revised pages into the applicable licensing document controlled copies. At a minimum, affected personnel must be briefed or trained on the revised requirements. Implementation can be complex and may be linked to other plant processes, such as refueling outages and plant modifications. It is necessary to ensure that the utility will be in compliance with the revised licensing document when it is implemented.
Process
a. The organizational responsibility and timing for identifying change impacts, as well as the process for tracking the activities to closure, is utility-specific.
b. Special consideration should be given to identification and scheduling of actions that must be performed prior to implementation of a TS change (such as satisfying new or revised Surveillance Requirements).
c. Identification of change impacts may be performed prior to or after submittal of the [[LAR|LAR]] to the NRC. However, sufficient consideration of change impacts should be given when drafting the [[LAR|LAR]] to determine any effect on the requested implementation period (e.g., is an outage needed to implement the change).
d. The [[RA|RA]] organization should keep the organizations responsible for making implementation changes informed of the [[LAR|LAR]] review progress.

==7 PEER REVIEW==
Objective
To confirm the [[LAR|LAR]] encompasses all necessary regulatory requirements and that the [[LAR|LAR]] and associated verification information is complete and accurate.
Background
Independent peer review of a draft [[LAR|LAR]] within the [[RA|RA]] organization ensures that the document meets quality expectations and regulatory requirements prior to distribution of the document to the organization for review.

Process
a. [[RA|RA]] management should assign a peer reviewer who is knowledgeable of the [[LAR|LAR]] requirements and the subject matter.
b. The peer reviewer should evaluate the document for format, content, presentation, and completeness, confirm the identification of verifiable statements, and verify the objective evidence supports the verifiable statements.
c. For Type 1 [[LAR|LAR]]s, the peer reviewer should confirm that all regulatory requirements and guidance is identified and followed. The peer reviewer should review the [[LAR|LAR]] against the acceptance criteria in NRC [[NRC LIC::LIC-109|LIC-109]].
d. For Type 2 and 3 [[LAR|LAR]]s, the peer reviewer should confirm the model application is followed and that that there are no significant technical deviations from the NRC-approved generic justification, model application, and, if applicable, model Safety Evaluation. The Peer Review should verify that any deviations from the model are clearly explained and justified.
e. The peer reviewer should confirm the discussion of plant-specific differences and verification of information is accurate and complete.
f. The peer reviewer should confirm that the precedent information, such as Requests for Additional Information on similar [[LAR|LAR]] submittals for other plants, have been adequately addressed.

==8 TECHNICAL REVIEW==
Objective
To ensure the technical content in the [[LAR|LAR]] is complete and accurate.
Background
The [[RA|RA]] preparer typically works with the responsible technical organizations while preparing a [[LAR|LAR]]. In creating the [[LAR|LAR]] narrative, the preparer may reorganize and reword technical organization input into a form suitable for the application. It's important for the responsible technical organization to verify that the technical content as presented in the [[LAR|LAR]] is complete and accurate and supports the conclusions in the [[LAR|LAR]].
Process
a. The [[RA|RA]] organization coordinates the [[LAR|LAR]] review by the appropriate technical organizations, based on the subject/technical content of the [[LAR|LAR]] (e.g., Operations, Engineering, Security, Emergency Planning). [[RA|RA]] is responsible to ensure that all information is reviewed by the responsible organizations.

b. The technical organization verifies that the [[LAR|LAR]] content is technically correct and is presented in the appropriate context, and accurately supports the conclusions in the [[LAR|LAR]].
c. The responsible technical organizations validate the accuracy of each verifiable statement and confirms that either objective evidence supports the statement or engineering judgment is supported by a knowledgeable independent technical reviewer.
d. The utility-specific process should ensure that significant changes to the [[LAR|LAR]] that are made as a result of the review process are subject to the same review and concurrence steps as the original [[LAR|LAR]], when appropriate.

==9 CONCURRENCE REVIEW==
Objective
To ensure that utility organizations that provided input to the [[LAR|LAR]] concur with the [[LAR|LAR]] content. Management responsible for the [[LAR|LAR]], that are affected by the [[LAR|LAR]], and whose organizations have provided input to the [[LAR|LAR]] should confirm that the [[LAR|LAR]] is consistent with company goals and objectives.
Background
Development of a [[LAR|LAR]], NRC review, and implementation can be resource-intensive activities and a limited number of [[LAR|LAR]]s can be developed and reviewed by the utility and the NRC. Management concurrence with a proposed [[LAR|LAR]] is a business necessity to ensure utility resources are being expended in a manner consistent with company goals and objectives.
Management concurrence with the [[LAR|LAR]] is a responsibility distinct from [[RA|RA]] or technical organization's management approval of their organization's review of the [[LAR|LAR]]. Concurrence with the [[LAR|LAR]] and acceptance of their organization's review of the [[LAR|LAR]] may be performed concurrently if the distinction is preserved in the utility-specific process (e.g., forms or checklists).
Process
a. Utility organizations that provided input to the [[LAR|LAR]] and the appropriate management, determined by the utility-specific process, is asked to concur with the [[LAR|LAR]].
b. Concurrence reviews should conclude:
* The [[LAR|LAR]] content correctly reflects the information provided during the development of the [[LAR|LAR]].
* The proposed change is consistent with the site and company goals and objectives,
* The [[LAR|LAR]] effectively addresses the need for change, and
* The benefit of the proposed change is commensurate with the expected cost.

==10 SAFETY COMMITTEE REVIEW==
Objective
To confirm the proposed change does not have a detrimental effect on nuclear safety.
Background
The utility-specific quality assurance ([[QA|QA]]) program or plant procedures typically require an onsite safety review committee. This committee is known by many different utility-specific names. The onsite safety review committee is typically required to review proposed [[LAR|LAR]]s to confirm there are no unacceptable effects on plant and public safety, security, and the environment prior to submittal to the NRC for review.
The [[QA|QA]] program or plant procedures may require an offsite safety review committee, nuclear safety review board, or other organization to provide an outside perspective on plant operation. This organization may be required to review [[LAR|LAR]]s before or after submittal to the NRC. In some cases, an independent reviewer is required to review a [[LAR|LAR]] before or after submittal to the NRC.
Process
a. The onsite safety review committee should review the [[LAR|LAR]] in accordance with the utility-specific process.
b. For Type 1 [[LAR|LAR]]s, the [[RA|RA]] Manager or other utility management should have the flexibility to determine if offsite safety review committee or independent review should occur before or after submittal of the [[LAR|LAR]] to the NRC for review, based on any new or unique characteristics of the [[LAR|LAR]].
c. If [[LAR|LAR]] review is required by the offsite safety review committee or an independent reviewer by the [[QA|QA]] program or plant procedures, the review of Type 2 and 3 [[LAR|LAR]]s should be excluded or should occur after submittal to the NRC for review.

==11 APPROVAL AND SUBMITTAL==
Objective
To approve and submit the [[LAR|LAR]] for NRC review in accordance with [[CFR::10 CFR 50.90#|10 CFR 50.90]] and [[CFR::10 CFR 50.30#b|10 CFR 50.30(b)]].

Background
In accordance with [[CFR::10 CFR 50.30#b|10 CFR 50.30(b)]], each amendment request must be executed in a signed original by a duly authorized officer under oath or affirmation.
Official licensing documents were historically submitted under "Oath or Affirmation" and notarized by a licensed notary. The law was changed in 1976 to allow an unsworn declaration. 28 U.S. Code § 1746," Unsworn declarations under penalty of perjury," states:
Wherever, under any law of the United States or under any rule, regulation, order, or requirement made pursuant to law, any matter is required or permitted to be supported, evidenced, established, or proved by the sworn declaration, verification, certificate, statement, oath, or affidavit, in writing of the person making the same (other than a deposition, or an oath of office, or an oath required to be taken before a specified official other than a notary public), such matter may, with like force and effect, be supported, evidenced, established, or proved by the unsworn declaration, certificate, verification, or statement, in writing of such person which is subscribed by him, as true under penalty of perjury, and dated, in substantially the following form:
...
(2) If executed within the United States, its territories, possessions, or commonwealths: "I declare (or certify, verify, or state) under penalty of perjury that the foregoing is true and correct. Executed on (date).
(Signature)
The [[LAR|LAR]] standard format described in Appendix A contains provisions for a notarized signature or an unsworn declaration.
Process
a. The [[LAR|LAR]] is approved by the signature authority and submitted to the NRC in accordance with legal requirements and utility-specific processes. A copy of the [[LAR|LAR]] must be sent to the state in accordance with [[CFR::10 CFR 50.91#b1|10 CFR 50.91(b)(1)]]. It is recommended that a courtesy copy of the [[LAR|LAR]] be provided to the NRC Project Manager ([[PM|PM]]) and that the utility inform the [[PM|PM]] that the [[LAR|LAR]] has been submitted.
b. The use of electronic submittal of [[LAR|LAR]]s is recommended as it normally allows the submittal to be available in the NRC's [[ADAMS|ADAMS]] more quickly than a physical submission. The email confirming submittal of the [[LAR|LAR]] should be retained until the document appears in [[ADAMS|ADAMS]]. Documents can also be submitted on Compact Disk (CD), but file format restrictions apply. For more information, see the NRC document, "Guidance for Electronic Submissions to the NRC" and the "Electronic Submittals" section of the NRC's website (http://www.nrc.gov/site-help/e-submittals.html).
c. If electronic submission is not used, the licensee should contact the NRC [[PM|PM]] to verify NRC receipt and docketing of the [[LAR|LAR]].

==12 NRC ACCEPTANCE REVIEW==
Objective
To ensure the NRC staff accepts the [[LAR|LAR]] for detailed technical review. This may require the utility to respond in a timely manner to any NRC requests for supplemental information necessary to make the [[LAR|LAR]] acceptable for review.
Background
The NRC staff conducts an "acceptance review" of incoming [[LAR|LAR]]s in accordance with [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-109|LIC-109]]. The main steps in the acceptance review process are:
* The NRC receives the [[LAR|LAR]] (the receipt date is the date the [[LAR|LAR]] becomes publicly available in [[ADAMS|ADAMS]]).
* NRC establishes an internal acceptance review schedule and assigns staff resources.
* NRC reviews the [[LAR|LAR]] for administrative and technical sufficiency.
* The NRC documents the results of the acceptance review in a letter or e-mail to the licensee.
An acceptance review has three possible results:
* Acceptable for Review – In this case, the NRC [[PM|PM]] notifies the contact identified in the submittal letter and initiates the technical review.
* Unacceptable with Opportunity to Supplement – In this case, the NRC [[PM|PM]] contacts the applicant to arrange a telephone call to discuss shortcomings in the submittal. During the call, the participants discuss an appropriate course of action for supplementing the initial submittal. The information must be submitted within the agreed-upon time frame, or the NRC may issue a letter of non-acceptance.
* Unacceptable with No Opportunity to Supplement – In this case, the NRC staff and management conclude the deficiencies are too significant to complete the acceptance review and decline to accept the submittal for review. The licensee may withdraw or revise and resubmit the [[LAR|LAR]].
If the licensee is given an opportunity to supplement the [[LAR|LAR]], the supplemental information is required to be provided quickly (e.g. typically less than 30 days). See [[NRC LIC::LIC-109|LIC-109]] for the timeline. If there is insufficient time to supplement the [[LAR|LAR]], the licensee may withdraw the [[LAR|LAR]].
Process
a. Submittal of supplemental information should follow, to the greatest extent possible given the scope of the information and the requested schedule, the [[LAR|LAR]] process described above.
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[[RA|RA]] management or other utility management may omit, perform after submittal, or alter some steps depending on the nature of the information being provided to the NRC.
b. If discussions with the NRC or informal or formal requests for supplemental information are inconsistent with the model application for a Type 2 or 3 [[LAR|LAR]], contact the Technical Specifications Task Force or the author of the NRC approved model application for support.

==13 NRC REQUESTS [[FOR|FOR]] ADDITIONAL INFORMATION PROCESS AND AUDITS==
Objective
To gain NRC staff approval of the [[LAR|LAR]] by responding in a timely manner to any NRC requests for additional information ([[RAI|RAI]]s) necessary to issue the amendment to the utility.
Background
The NRC may issue one or more [[RAI|RAI]]s on a [[LAR|LAR]]. Each [[RAI|RAI]] may contain one or more questions that the NRC uses to address gaps in the safety evaluation. The NRC uses the [[RAI|RAI]] process when information believed to be necessary to complete the NRC safety evaluation is not contained in the [[LAR|LAR]] or in any other docketed correspondence, or cannot reasonably be inferred from other sources of information readily accessible by the NRC staff. Frequent and early communication between the [[PM|PM]], technical staff, and the licensee can minimize the need for RAls.
Draft [[RAI|RAI]]s are typically provided electronically to allow the utility to review the questions and request clarifications. A teleconference may be scheduled to discuss the draft questions to ensure the final questions are clear. In some cases, the utility may volunteer to respond to an informally transmitted [[RAI|RAI]]. The licensee always has the option to ask the NRC to send an [[RAI|RAI]] formally.
The NRC may also initiate informal communication to allow the technical staff to discuss background information and clarify understanding. This information does not form part of the basis for approval, but must still be complete and accurate.
An NRC audit can be an effective method of sharing complex or voluminous information with the NRC in lieu of multiple [[RAI|RAI]]s. It is also a means to understand the [[RAI|RAI]] and work with the NRC to ensure the proper information will be provided in the [[RAI|RAI]] response. NRC guidance for the conduct of an audit is found in [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-111|LIC-111]], "Regulatory Audits."
Process
a. Prior to issuance of the formal [[RAI|RAI]] letter by the NRC, the NRC will typically provide a draft copy of the [[RAI|RAI]] and provide the opportunity for a conference call to obtain clarification of the draft [[RAI|RAI]] questions.

b. If discussions with the NRC or informal or formal [[RAI|RAI]]s are inconsistent with the model application for a Type 2 or 3 [[LAR|LAR]], contact the Technical Specifications Task Force or the author of the NRC approved model application for support.
c. Use a clear format to respond to each question in the [[RAI|RAI]]. A recommended format is to repeat each question from the NRC and then provide the answer.
d. Response to an [[RAI|RAI]] should follow, to the greatest extent possible given the scope of the requested information, the [[LAR|LAR]] process described above. The [[RA|RA]] management or other utility management may omit, perform after submittal, or alter some steps depending on the nature of the information.
e. There should be a statement, usually in the [[RAI|RAI]] response cover letter, that the No Significant Hazards Consideration ([[NSHC|NSHC]]) analysis provided in the original submittal is or is not altered by the additional information provided by the licensee. If the [[NSHC|NSHC]] analysis is affected by the additional information, then a revised or new [[NSHC|NSHC]] must be included in the [[RAI|RAI]] response. A revised [[NSHC|NSHC]] analysis may require the NRC to republish the revised analysis in the Federal Register.
f. The utility should consider, based on the complexity of the [[RAI|RAI]] questions and the utility responses, sharing the [[RAI|RAI]] responses with the NRC in draft to ensure the questions were adequately addressed. A teleconference or meeting to discuss the draft responses may avoid a second-round [[RAI|RAI]].
g. Response to the [[RAI|RAI]] should be within the schedule described in the [[RAI|RAI]]. Note that [[CFR::10 CFR 2.108#|10 CFR 2.108]] states, "(a) The Director of Nuclear Reactor Regulation ... may deny an application if an applicant fails to respond to a request for additional information within thirty (30) days from the date of the request, or within such other time as may be specified."

==14 NRC APPROVAL, REJECTION, [[OR|OR]] WITHDRAWAL==
Objective
To prepare the utility to implement an approved [[LAR|LAR]] within the implementation period, or to assure the affected utility organizations are informed of withdrawal or NRC rejection of a [[LAR|LAR]].
Background
A [[LAR|LAR]] may be approved or rejected by NRC, or withdrawn by the licensee. Approved amendments are implemented by the licensee. Rejected amendments may be appealed by the licensee, or resubmitted in revised form. Withdrawn amendments may be abandoned or resubmitted at some future time.
The NRC documents the basis of their decision in a safety evaluation ([[SE|SE]]). Licensees should review the [[SE|SE]] against the [[LAR|LAR]] and any [[RAI|RAI]] supplements for accuracy. If significant errors are found, request the NRC to revise the [[SE|SE]] to correct the errors.
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License amendments are effective immediately on approval by the NRC. However, the NRC typically provides a limited period for the licensee to implement the change (e.g., revise procedures, perform required tests or modifications, perform training, etc.) There is no regulatory requirement for this temporary allowance permitting use of the former license requirements for a specified period before imposing the new requirements. However, the implementation period is specified in the license amendment and any change to the implementation period after the [[LAR|LAR]] is approved requires a separate [[LAR|LAR]].
Process
a. Contact the NRC [[PM|PM]] and request the opportunity to review the draft [[SE|SE]] to verify its consistency with the [[LAR|LAR]] and any [[RAI|RAI]] responses.
b. Prior to issuance of the amendment, verify there are no conflicts with recently issued amendments or [[LAR|LAR]]s under NRC review, such as common TS pages or amendment numbers.
c. Review the issued [[SE|SE]] against the [[LAR|LAR]], as modified by any [[RAI|RAI]] responses, and verify the [[SE|SE]] accurately reflects the proposed change. If not, inform the NRC [[PM|PM]]. For errors that conflict with the amendment request, the licensee should request a revised [[SE|SE]] from the NRC.
d. On approval of the [[LAR|LAR]], [[RA|RA]] should inform the organizations responsible for any implementation actions of the implementation date and track, by utility-specific methods, the actions.
e. If a [[LAR|LAR]] is rejected or withdrawn, [[RA|RA]] should inform the responsible organization that implementation actions are not required and identify any additional actions needed in lieu of the approved [[LAR|LAR]].
f. [[RA|RA]] should initiate the utility-specific actions to distribute the change to the document copy holders within the required implementation period.

==15 IMPLEMENTATION==
Objective
To ensure the issued License Amendment is implemented within the required time frame and the new requirements are met.
Background
Implementation of a license amendment includes many activities, such as updating the applicable controlled copies, training or briefing key stakeholders (e.g., operators), installing modifications, performing new or revised Surveillance Requirements, revising procedures, programs, and training, and revising the TS Bases. Implementation activities necessary to comply with the revised license requirements must be implemented within the implementation period specified in the license amendment or an exception must be included in the amendment.
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Except in the unusual case in which the Bases changes are the sole focus of the [[LAR|LAR]] (because the [[CFR::10 CFR 50.59#|10 CFR 50.59]] review of the Bases changes determined that prior NRC approval was required AND there are no TS, [[OL|OL]], or [[UFSAR|UFSAR]] changes that need to be made in parallel), the NRC does not typically approve the Bases changes as part of the [[LAR|LAR]] approval. Instead, the NRC approves the technical change and the revised Bases pages are issued by the licensee under the Technical Specification Bases Control Program as part of the implementation of the amendment.
If the information being included in the Bases is consistent with the information in the "Technical Evaluation" or the "Regulatory Evaluation" sections of the [[LAR|LAR]] to the extent that the Bases information can be considered to be "directly related" to the [[LAR|LAR]] once it is approved by the NRC, the implementation of the Bases change under the Technical Specification Bases Control Program is simplified. This is because, as noted in Section 4.1.1 of [[NEI::NEI 96-07|NEI 96-07]], "Guidelines for [[CFR::10 CFR 50.59#|10 CFR 50.59]] Implementation," only activities (such as the Bases change) that are not "directly related" to the required TS change are subject to [[CFR::10 CFR 50.59#|10 CFR 50.59]]. This is justifiable because activities that are directly related to the amendment are encompassed in the NRC review that has already been completed, and the purpose of [[CFR::10 CFR 50.59#|10 CFR 50.59]] is to determine if an activity needs to be approved by the NRC.
Therefore, if the Bases changes are directly related to the license amendment, they are simply a follow-on activity and can be implemented without the need for a [[CFR::10 CFR 50.59#|10 CFR 50.59]] review. Conversely, if there are aspects of the proposed Bases changes that are not directly related to the approved license amendment, then those aspects are subject to review under [[CFR::10 CFR 50.59#|10 CFR 50.59]] prior to implementation.
Process
a. The responsible organizations perform the implementation actions. Tracking of the implementation actions is a utility-specific process.
b. The responsible organizations inform [[RA|RA]] when any actions that must be performed before or concurrent with implementing the amendment are complete. If there is any potential delay in completing the actions within the implementation period for the [[LAR|LAR]], [[RA|RA]] should be informed.
c. TS Bases changes associated with approved TS changes are reviewed, approved, and implemented in accordance with the utility-specific implementation of the Technical Specifications Bases Control Program.
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=APPENDIX A STANDARD FORMAT [[FOR|FOR]] PLANT-SPECIFIC LICENSE AMENDMENT REQUESTS=
This Appendix provides a standardized format that licensees may use on a voluntary basis to request approval of a Type 1 (i.e., plant-specific) [[LAR|LAR]]. Information in brackets represents amendment-specific information to be inserted by the licensee. Footnotes are used to explain certain concepts. Thus, they are part of this guideline, not part of the [[LAR|LAR]] format.
Licensees are encouraged to follow the guidance in this Appendix, including order, titles, and level of detail. However, document formatting, such as title location, pagination, use of emphasis (e.g., bold, underline, etc.), are left to the licensee's preference.
A successful [[LAR|LAR]] is not a collection of facts. A successful [[LAR|LAR]] is a carefully crafted narrative that compellingly describes the current condition, the need for change, the proposed change, and the technical justification for the change, while meeting the regulatory and administrative requirements. It is recommended that licensees review draft [[LAR|LAR]]s not only for accuracy, but for continuity and readability.
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COVER LETTER
[Licensee’s letterhead]
[Date] [[CFR::10 CFR 50.90#|10 CFR 50.90]]
ATTN: Document Control Desk
U.S. Nuclear Regulatory Commission
Washington, [[DC|DC]] 20555-0001
[Plant / Unit Name(s)]
[Docket No(s) [50-___, 50-___]
Subject: [Title of the proposed license amendment request]3
Pursuant to [[CFR::10 CFR 50.90#|10 CFR 50.90,]] [license holder] hereby requests [include a brief summary of the proposed amendment and the results of the corresponding "no significant hazards consideration determination." [If the proposed amendment is based on a Technical Specification Task Force ([[TSTF|TSTF]]) traveler that does not have an NRC-accepted model application, include a statement to that effect including the [[TSTF|TSTF]] traveler number (TSTF-xxx) and title.].
Approval of the proposed amendment is requested by [date + justification]4. Once approved, the amendment shall be implemented within [ ] days.5
[If the [[LAR|LAR]] is non-voluntary, see the guidance in Appendix G and consider adding a discussion regarding the non-voluntary nature of the [[LAR|LAR]].]
[If regulatory commitments are made in the submittal, include here (and in an attachment to the Enclosure) a listing of the formal licensee commitments that would apply when NRC approves the amendment. If no regulatory commitments are made, include a statement to that effect in the cover letter.]
[In accordance with [[CFR::10 CFR 50.91#|10 CFR 50.91,]] [name of licensee] is notifying the State of [name of state] of this [[LAR|LAR]] by transmitting a copy of this letter and enclosure to the designated State Official.]
[If there are any questions or if additional information is needed, please contact [licensee’s point of contact for the NRC Office of Nuclear Reactor Regulation] at [telephone number and/or e-mail address].]
3 [The title used by many licensees is "License Amendment Request ([[LAR|LAR]])." Other licensees use "Proposed License Amendment (PLA)." These and other equivalent terms are acceptable titles. See Appendix G regarding the title of non-voluntary [[LAR|LAR]]s.]
4 [Provide justification in the cover letter for the "need date." For example, if approval by a certain date is necessary to prepare for startup after a refueling outage.]
5 [A 60-120 day implementation period is typical. If additional implementation time is needed, provide justification in the cover letter, e.g., if significant procedure changes are necessary to support implementation, or if significant plant modifications require a refueling outage for installation.]
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[In accordance with [[CFR::10 CFR 50.30#b|10 CFR 50.30(b)]], a license amendment request must be executed in a signed original under oath or affirmation. This can be accomplished by including a notarized affidavit confirming the signature authority of the signatory, or by including the following statement in the cover letter: "I declare under penalty of perjury that the foregoing is true and correct. Executed on (date)." The alternative statement is pursuant to 28 USC 1746. It does not require notarization, but must be quoted verbatim as written here.]
Sincerely, [this closing is optional if the preceding 28 USC 1746 statement is used].
[Signature]
[Name]
[Title, if not already included in the letterhead]
Enclosure: Evaluation of the Proposed Change
cc: [NRC Region _] [required by [[CFR::10 CFR 50#|10 CFR 50]].4]
[NRC Project Manager] [optional, but recommended]
[NRC Resident Inspector(s)] [required by [[CFR::10 CFR 50#|10 CFR 50]].4]
[State of ________] [required by [[CFR::10 CFR 50#|10 CFR 50]].4]
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ENCLOSURE
Evaluation of the Proposed Change
A table of contents with page numbers is optional, but is recommended for large [[LAR|LAR]]s.
Subject: Brief title. Should be consistent with the cover letter subject.]
1. SUMMARY DESCRIPTION
2. DETAILED DESCRIPTION
2.1 System Design and Operation
2.2 Current Technical Specifications Requirements
2.3 Reason for the Proposed Change
2.4 Description of the Proposed Change
3. TECHNICAL EVALUATION
4. REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements/Criteria
4.2 Precedent [optional]
4.3 No Significant Hazards Consideration Analysis
4.4 Conclusions
5. ENVIRONMENTAL CONSIDERATION
6. REFERENCES
-------------------------------
ATTACHMENTS:6
1. List of Regulatory Commitments [If Needed]
2. Technical Specification (and/or Operating License) Page Markups [required]
3. Bases Page Markups [optional, see discussion in Appendix A]
4. Retyped Technical Specification (and/or Operating License) Pages [see discussion in Appendix A]7
6 [Number the attachments based on which optional attachments are included with the Enclosure.]
7 [Retyped or "camera ready" must be provided to support the license amendment request. The pages may be included in the original submittal or provided at the end of the review process to accommodate revisions derived from responses to NRC Requests for Additional Information or other sources. See discussion in Enclosure Attachment 4.]
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1 SUMMARY DESCRIPTION
The summary description should be a brief (1-2 sentence) description of the proposed change written for the public. It should be consistent with the description of the change in the cover letter and in the introduction of the No Significant Hazards Determination Analysis. The summary description should also be suitable for the NRC to use in the introduction of the safety evaluation for the change.
2 DETAILED DESCRIPTION
2.1 System Design and Operation
Describe the systems, structures, or components ([[SSC|SSC]]s) affected by the proposed change. Describe the system operation at a level of detail appropriate for someone knowledgeable with nuclear technology but not familiar with the particular nuclear steam supply system ([[NSSS|NSSS]]) or plant design. Highlight any unique design features relevant to the change.
Do not include irrelevant information regarding the system, such as vents and drains, secondary system uses, etc. Only include information needed to understand the proposed change.
If it is relevant to the proposed change, discuss how the [[SSC|SSC]] responds in analyzed accidents.
2.2 Current Technical Specifications Requirements
Describe the current TS requirements that are relevant to the change. Discuss the LCO Bases description of an operable [[SSC|SSC]], if it's related to the change. Discuss any relevant Surveillance Requirements.
For plants with TS based on the Standard Technical Specifications ([[STS|STS]]), describe any relevant differences between the plant TS and the current [[STS|STS]].
For plants with TS not based on the [[STS|STS]], describe any relevant differences between the plant TS requirements and the [[STS|STS]]. Licensees should pay particular attention to requirements in other TS that may affect the proposed change, such as diesel generator requirements in TS on supported [[SSC|SSC]]s.
The intent is that Sections 2.1 and 2.2 will provide the NRC reviewer an adequate understanding of the relevant system design and TS requirements to review the proposed change.
2.3 Reason for the Proposed Change
Explain the reason why the license amendment is being requested. The length of the section may vary. Licensees should take caution when describing the reasons for the TS or [[OL|OL]] change to not imply NRC review and approval of related licensee actions, such as design changes, is also requested.
If the [[LAR|LAR]] is non-voluntary, see the guidance in Appendix G. If appropriate, include a discussion such as "This license amendment request ([[LAR|LAR]]) is required to comply with [new or
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amended regulation.] This [[LAR|LAR]] is not a voluntary request from a licensee to change its licensing basis and it is not subject to 'forward fit' considerations as described in the letter from S. Burns (NRC) to E. Ginsberg (NEI), dated July 14, 2010 ([[ADAMS|ADAMS]] Accession Number [[document::ML01960180|ML01960180]])."]
2.4 Description of the Proposed Change
Describe the proposed change to the TS or [[OL|OL]] as succinctly and clearly as possible. A red-line/strikeout of the proposed changes is recommended. The intent this section is to unambiguously show the change to the TS or [[OL|OL]] being requested, not to explain or justify the change.
3 TECHNICAL EVALUATION
The Technical Evaluation section should be a logical continuation of the narrative begun in the Detailed Description section. Information in the Detailed Description section should not be repeated unless needed for clarity. If there is a large discussion that would interrupt the narrative, such as a multiple page table or a detailed calculation description, consider placing the material in an attachment and only present the conclusion in the body. The explanation should be stated simply, avoiding long or unnecessary sentences.
A logical order should be used to present the evaluation of the change. For [[LAR|LAR]]s that change multiple requirements (LCO, SRs, Actions, etc.), it is typically best to present the justification in the same order used in the "Proposed Change" section.
The Technical Evaluation should demonstrate an adequate level of safety for the change, including, if applicable, analytical methods and input parameters. Describe why the method is acceptable and whether it is in the current licensing basis or is an NRC-approved method.
If information from the [[UFSAR|UFSAR]] is needed to support the change, paraphrase or quote the information. NRC reviewers may not have direct access to plant-specific documentation.
The Technical Evaluation section should include all information specifically required by guidance documents for the type of submittal, such as [[NRR|NRR]] office instructions and Regulatory Guides. Ensure that all acceptance review criteria are met. Ensure that all referenced documents, such as travelers and topical reports, are followed and any limitations are met. Describe any issues in precedent submittals and how they are addressed.
When appropriate, compare the plant's current licensing basis and the proposed change to the current version of the Standard Review Plan ([[SRP|SRP]]) ([[NUREG::NUREG-0800|NUREG-0800]]). Justify any deviations from the [[SRP|SRP]].
Ensure that all proposed changes in the [[LAR|LAR]] are evaluated, including administrative or editorial corrections, such as elimination of expired one-time allowances.
If the proposed amendment is risk-informed, see the guidance in Appendix E.
If TS Bases are included in the [[LAR|LAR]] and marked "for information only," the changes to the Bases should be briefly discussed in the "Technical Evaluation" Section of the [[LAR|LAR]] to the extent
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necessary to ensure that the Bases information will be considered "directly related" to the [[LAR|LAR]] once it is approved by the NRC. This eases processing of the Bases change under the Technical Specifications Bases Control Program because as noted in Section 4.1.1 of [[NEI::NEI 96-07|NEI 96-07]], "Guidelines for [[CFR::10 CFR 50.59#|10 CFR 50.59]] Implementation," only activities (such as the Bases changes) that are not "directly related" to the required TS change are subject to [[CFR::10 CFR 50.59#|10 CFR 50.59]].
Briefly summarize the arguments at the end of the section to assist the NRC staff in writing the Safety Evaluation. The Technical Evaluation section should be written such that excerpts may be extracted for use in the NRC staff’s [[SE|SE]].
4 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements/Criteria
The Regulatory Evaluation provides a basis that the NRC staff may use to find the proposed amendment acceptable by describing how the proposed change satisfies the applicable regulatory requirements and criteria. It should be written such that excerpts may be used in the NRC staff’s [[SE|SE]].
It is recommended that a list or table of applicable regulatory requirements or criteria be included. The section should conclude with a statement similar to, "The proposed change does not affect compliance with these regulations or guidance and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met."
It is recommended that a description of the plant's licensing basis related to General Design Criteria (GDC) be included if the [[LAR|LAR]] is not administrative.
4.2 Precedent [Optional]
Effective evaluation and presentation of precedent-setting licensing actions can reduce [[LAR|LAR]] preparation efforts and improve the overall quality of the application, minimize NRC [[RAI|RAI]]s, and improve the efficiency of the regulatory review process. Precedent, by itself, does not demonstrate the acceptability of a proposed amendment. The citation of precedent by a licensee in a [[LAR|LAR]] is voluntary, and it should be used only to the extent that it supports the review.
It is important to distinguish licensing precedent from other regulatory or technical considerations relevant to the requested licensing action. For example, a vendor topical report or [[TSTF|TSTF]] traveler, even when evaluated and approved in an NRC Safety Evaluation, is technically not a licensing action, and should not be identified as precedent (but may be used in the Technical Evaluation).
Precedent may be identified through various sources, such as the NRC [[ADAMS|ADAMS]], the Federal Register, or commercial licensing information services.
There are several considerations in evaluating the use of precedent in [[LAR|LAR]]s. The licensee must determine the extent to which potential precedent is similar and relevant to the proposed action. Similarities and differences between the precedent and proposed actions must be evaluated to determine the effect on the applicability of the precedent to the proposed change. The NRC staff
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uses precedent to make reviews more efficient but is not controlled by precedent when reviewing a [[LAR|LAR]]. For example, a change may require greater justification than a precedent action if the regulatory or design margins are smaller or the uncertainties are larger than the precedent action, or if the NRC staff has questions on the use of engineering judgment.
The determination of relevance to the proposed action includes a comparison of the actual content of the precedent, including the original [[LAR|LAR]], any supplements, [[RAI|RAI]]s and responses, and the NRC [[SE|SE]], and the proposed change. The preparer should also understand similarities and differences in the design and operation of systems, structures, and components ([[SSC|SSC]]s). Differences in wording, grammar, punctuation and structure, especially when changes to TS are involved, should be closely evaluated to ensure that any editorial differences do not also result in technical differences. If differences are extensive, the citation of precedent in the application should be reconsidered. Citing a precedent that requires extensive justification of the differences will likely hinder the review. The NRC has noted that citing recent precedent [[LAR|LAR]]s may assist the NRC [[PM|PM]] in requesting a reviewer that is familiar with the relevant issues.
The precedent citation in the [[LAR|LAR]] should identify the affected licensee, power plant and amendment number. References to related documents and [[ADAMS|ADAMS]] Accession Numbers., (e.g. [[LAR|LAR]]s, [[LAR|LAR]] supplements, [[RAI|RAI]]s and responses), should be provided as necessary to support the above discussion. Include a brief discussion of how previous NRC considerations and decisions constitute precedent for the proposed licensing action. Similarities and differences between the precedent licensing action and the proposed amendment should be identified. Additionally, relevant plant-specific similarities and differences, including those in plant design and licensing basis, should be described. The effect of the similarities and differences should be discussed both to describe the differences between the precedent and the proposed actions, and to point out any limitations on the relevance of the precedent action.
NRC staff guidance for the consideration of precedent in [[LAR|LAR]] is in [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-101|LIC-101]], "License Amendment Review Procedures." It states that precedent is intended to "enhance NRR’s efficiency in responding to the needs of both the licensees and the public." Effective consideration of licensing precedent supports the following specific objectives of LIC-101:
* Promote consistency in processing of license amendments, and
* Increase technical consistency similar licensing actions
The NRC staff reviews proposed precedent for applicability, accuracy, and completeness when compared with the incoming [[LAR|LAR]] and its associated plant-specific design details. The staff verifies that the precedent is appropriate for use with the [[LAR|LAR]] and that it meets current NRC expectations with respect to format, content, guidance, and conclusions.
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4.3 No Significant Hazards Consideration Determination Analysis8
Provide a brief summary description of the proposed change written for the public. It should be consistent with the description in Section 1, "Summary Description." Redefine any [[Acronym|acronym]]s and avoid the use of technical jargon. Note that in the analysis, the entire [[LAR|LAR]] is a single "proposed change."
The purpose of the No Significant Hazards Consideration Determination analysis is to determine if a requested public hearing on the [[LAR|LAR]] should be held before or after issuance of the amendment. The NSHCD analysis does not determine if a change is safe or acceptable.
The NSHCD Analysis should not include any proprietary information and should not include specific values or parameters. The NSHCD is published in the Federal Register early in the review of a [[LAR|LAR]]. If a supplement to the [[LAR|LAR]] changes information in the Federal Register Notice, a revised notice must be published and the public comment period is restarted.
Consult the NRC [[Regulatory Issue Summary::RIS 2001-22|Regulatory Issue Summary 2001-22]], "Attributes of a Proposed No Significant Hazards Consideration Determination," for guidance. Typically, one or two paragraphs per criterion are sufficient.
Do not include new concepts or arguments in the NSHCD Analysis that are not discussed in the justification for the [[LAR|LAR]].
The format of the NSHCD Analysis is typically similar to:
[Licensee name] has evaluated whether a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in [[CFR::10 CFR 50.92#|10 CFR 50.92,]] "Issuance of amendment," as discussed below:
1) Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
[For guidance on preparing a basis for this response, see the First Standard from [[Regulatory Issue Summary::RIS 2001-22|RIS 2001-22]] (Ref. 8: Consider the effect of the change on structures, systems, and components ([[SSC|SSC]]s) of the plant to determine how the proposed change affects plant operations, any design function or an analysis that verifies the capability of an [[SSC|SSC]] to perform a design function. Determine if the proposed amendment would change any of the previously evaluated accidents in the [[UFSAR|UFSAR]]. The word ‘accidents’ refers to anticipated (or abnormal) operational [[Transient|transient]]s and postulated design basis accidents, including the events with which the plant must be able to cope (e.g., earthquake, flooding, turbine missiles, and fire) as described in the [[UFSAR|UFSAR]]. Determine if [[SSC|SSC]]s,
8 [General guidance is contained in NRC [[Regulatory Issue Summary::RIS 2001-22|Regulatory Issue Summary 2001-22]], "Attributes of a Proposed No Significant Hazards Consideration."]
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operating procedures, and administrative controls that are affected have the function of preventing or mitigating any of these accidents. If the proposed change increases the likelihood of the malfunction of an [[SSC|SSC]], the potential impact on analyzed accidents should be considered (e.g., an increased likelihood of an [[SSC|SSC]] malfunction may increase the probability or consequences of an accident). If there is no impact on previously evaluated accidents, explain why.
Discuss the differences in the probability and consequences of these accidents (or the bounding scenario) before and after the change and whether the differences are significant. If the change is not considered significant, explain why. Whether an increase is significant should be assessed case-by-case. A qualitative judgment may need to be made. Values of probability or consequence that continue to meet the licensing basis or applicable guidelines in the Standard Review Plan are generally not considered significant changes. If the probability of occurrence remains within the ranges already presented in the [[UFSAR|UFSAR]] for initiating events, then the increase is not considered significant. An increase beyond any of these values that is not deemed significant should be justified. The significance determination should include a comparison of the value before the change to that after the change. A large increase might not be considered significant in one situation, but a relatively small increase might be significant in another situation. The licensee should adequately justify the proposed determination.]
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.
[For guidance on preparing a basis for this response, see the Second Standard from [[RIS|RIS]] 2001-22: Determine whether the proposed amendment will change the design function or operation of the [[SSC|SSC]]s involved, or whether interim processes (e.g., process of installing a new system component or construction of a new facility, performance of testing or maintenance) will affect the [[SSC|SSC]]s’ operation or its ability to perform its design function. Then determine whether the proposed change will create the possibility of a new or different kind of accident due to credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases. This new accident would have been considered a design basis accident in the [[UFSAR|UFSAR]] had it been previously identified. A new initiator of the same accident is not a different type of accident. Finally, the accident must be credible within the range of assumptions previously applied (e.g., random single failure, loss of off-site power, no reliance on non-safety-grade equipment).]
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
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3) Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
[For guidance on preparing a basis for this response, see the Third Standard from [[RIS|RIS]] 2001-22: Safety margins are applied at many levels to the design and licensing basis functions and to the controlling values of parameters to account for various uncertainties and to avoid exceeding regulatory or licensing limits. The specific values that define margin are established in each plant’s licensing basis. Licensees should identify the safety margins that may be affected by the proposed change and review the conservatism in the evaluation and analysis methods that are used to demonstrate compliance with regulatory and licensing requirements.
The safety margin before the change should be compared to the margin after the proposed change to determine if the amendment will reduce the margin, and if the change is significant. If a change does not exceed or alter a design basis or safety limit (i.e., the controlling numerical value for a parameter established in the [[UFSAR|UFSAR]] or the license) it does not significantly reduce the margin of safety. In other cases, the assessment of significance for this standard should be made on the same basis as discussed in the guidance for the first standard. Uncertainties and errors need to be considered in calculating the margin.]
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, [licensee name] concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in [[CFR::10 CFR 50.92#c|10 CFR 50.92(c)]], and, accordingly, a finding of "no significant hazards consideration" is justified.
4.4 Conclusions
In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission’s regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
5 ENVIRONMENTAL CONSIDERATION
The identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review is the subject of [[CFR::10 CFR 51.22#|10 CFR 51.22]] (Ref. 9). The categories of actions deemed "categorical exclusions" are specified by [[CFR::10 CFR 51.22#c|10 CFR 51.22(c)]]. The licensee’s consideration of environmental factors should include sufficient detail to support a finding of categorical exclusion. For most changes, it is clear that the environment will not be affected (e.g., extending a surveillance interval). One of the following two paragraphs will typically be applicable:
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A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in [[CFR::10 CFR 20#|10 CFR 20,]] or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in [[CFR::10 CFR 51.22#c9|10 CFR 51.22(c)(9)]]. Therefore, pursuant to [[CFR::10 CFR 51.22#b|10 CFR 51.22(b)]], no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
or
The proposed amendment is confined to (i) changes to surety, insurance, and/or indemnity requirements, or (ii) changes to recordkeeping, reporting, or administrative procedures or requirements. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in [[CFR::10 CFR 51.22#c10|10 CFR 51.22(c)(10)]]. Therefore, pursuant to [[CFR::10 CFR 51.22#b|10 CFR 51.22(b)]], no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
However, for the minority of instances where a proposed amendment does not qualify for a categorical exclusion (e.g., if special circumstances exist, or if the action does not meet applicable criteria in [[CFR::10 CFR 51.22#c|10 CFR 51.22(c)]]), the NRC staff will prepare an environmental assessment and may require the licensee to submit information in accordance with [[CFR::10 CFR 51.41#|10 CFR 51.41]].
6 REFERENCES
Identify and number references used in [[LAR|LAR]]. Each reference should be cited at least once in this Enclosure (Evaluation of the Proposed Change). If a reference is needed to understand, review, or approve the proposed amendment, it should be considered for inclusion as an attachment and identified with a suitable attachment number or letter.]
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ENCLOSURE
Attachment 1 [If Needed]
List of Regulatory Commitments
[Include this attachment if regulatory commitments are made in the submittal. If no regulatory commitments are made, include a statement to that effect in the cover letter.]
The following table identifies the regulatory commitments in this document. Any other statements in this submittal represent intended or planned actions. They are provided for information purposes and are not considered to be regulatory commitments.
COMMITMENT
TYPE
SCHEDULED COMPLETION DATE
One-Time
Continuing Compliance
[1. Duplicate the commitment wording from the body of the [[LAR|LAR]]. Guidance on controlling regulatory commitments is contained in [[NEI::NEI 99-04|NEI 99-04]] and [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-105|LIC-105]] (Ref.10)]
[2.]
[3.]
[4.]
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ENCLOSURE
Attachment 2
Operating License and/or Technical Specification Page Markups
Mark up affected Operating License and/or Technical Specification pages by either of the following methods:
* Word-processor mark-ups using the program’s "track changes " feature (recommended), or
* Hand-written mark-ups of copies of the affected pages
For plants with Improved Technical Specifications, the changes to the Technical Specifications should be consistent with TSTF-GG-05-01, "Writer's Guide for Plant-Specific Improved Technical Specifications," (Ref. 11). Avoid problem areas, such as:
* Using footnotes except in tables; and
* Not maintaining the page numbering. Do not insert additional pages by altering the page number format (e.g., page number "3.4.1-1A") and do not retain pages made blank by deletion of material. Reissue an entire specification or specification’s Bases if needed.
* Improperly formatting the Action and Surveillance tables starting and ending double lines.
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ENCLOSURE
Attachment 3
Technical Specifications Bases Page Markups
The NRC encourages licensees to include revised Bases pages with the license amendment request. For most license amendments, it is recommended that Bases changes be included to provide additional explanation to the NRC staff on how the licensee interprets and plans to implement the proposed change. The NRC has stated that providing the Bases pages contributes to the goal of increasing the quality of [[LAR|LAR]]s and may reduce confusion and the need for requests for [[RAI|RAI]]s.
Bases markups are typically marked "for information only" in the [[LAR|LAR]] as the NRC is not required to review and approve most Bases changes in accordance with [[CFR::10 CFR 50.90#|10 CFR 50.90]]. Instead, changes to the Bases are made by the licensee under the Technical Specifications Bases Control Program following NRC approval of the [[LAR|LAR]]. The Bases changes contribute to the NRC's understanding of the implementation of the proposed Technical Specification requirements. For example, Bases descriptions of what constitutes Operability of a system or Bases descriptions that provide details regarding how to perform risk assessments, Surveillance Requirements, or mitigating actions, would substantially contribute to the NRC's understanding of the proposed Technical Specifications and should be included for information in the [[LAR|LAR]]. Such pages should clearly be designated as "For Information Only" in the submittal, both in the cover letter and in the justification.
Mark up affected Technical Specification Bases pages by either of the following methods:
* Word-processor mark-ups using the program’s "track changes " feature (recommended), or
* Hand-written mark-ups of copies of the affected pages
For plants with Improved Technical Specifications, the changes to the Technical Specifications should be consistent with TSTF-GG-05-01, "Writer's Guide for Plant-Specific Improved Technical Specifications," (Ref. 12). Avoid problem areas, such as:
* Not maintaining the page numbering. Do not insert additional pages by altering the page number format (e.g., page number "3.4.1-1A") and do not retain pages made blank by deletion of material. Reissue an entire specification or specification’s Bases if needed.
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ENCLOSURE
Attachment 4
Retyped Operating License and/or Technical Specification Pages
Retyped or 'camera ready' pages must be provided to support the license amendment request. The pages may be included in the original submittal or provided at the end of the review process to accommodate revisions derived from responses to NRC Requests for Additional Information or other sources.
Providing retyped pages with the submittal should be considered for complex submittals to assist the NRC staff with their review and to potentially avoid [[RAI|RAI]]s.
Discuss with the NRC [[PM|PM]] whether to include retyped technical specification pages in the [[LAR|LAR]]. NRC staff can request retyped pages at any time if it needs this information to support the review. This request could come as early as the acceptance review if the staff determines that this information is needed to initiate the technical review.
Retyped pages should have revision bars indicating the affected sections. The page footer should have previous Amendment numbers removed (or clearly marked with a "strikethrough") and a "placeholder" for the new Amendment number inserted.
For plants with Improved Technical Specifications, the retyped changes to the Technical Specifications should be consistent with TSTF-GG-05-01, "Writer's Guide for Plant-Specific Improved Technical Specifications."
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=APPENDIX B [[LAR|LAR]] EXAMPLE=
This Appendix provides an example [[LAR|LAR]] based on the guidance in Appendix A. Licensees are encouraged to follow the guidance in this Appendix, including order, titles, and level of detail. However, document formatting, such as title location, pagination, use of emphasis (e.g., bold, underline, etc.), are left to the licensee's preference.
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COVER LETTER
DATE [[CFR::10 CFR 50.90#|10 CFR 50.90]]
ATTN: Document Control Desk
U.S. Nuclear Regulatory Commission
Washington, [[DC|DC]] 20555-0001
My-Plant Units 1 & 2
Docket Nos.[[SSC::07-50-001|50-001]] & 50-002
SUBJECT: License Amendment Request: Add Action for Two Inoperable CREATCS Trains
Pursuant to [[CFR::10 CFR 50.90#|10 CFR 50.90,]] My Power & Light hereby requests a license amendment to revise the Unit 1 and 2 Technical Specifications (TS). The proposed change will revise [[TS::TS-3.7.11|TS 3.7.11]], "Control Room Emergency Air Temperature Control System (CREATCS)," to add TS Actions for two inoperable CREATCS trains. The added Action provides 24 hours to restore a CREATCS train to operable status provided the control room area temperature is maintained below 90°F. The Action is modified by a Note which makes the Action not applicable if the condition is entered intentionally.
The Enclosure provides a description and assessment of the proposed changes. Attachment 1 contains a list of regulatory commitments associated with the requested change. Attachment 2 provides the existing TS pages marked to show the proposed changes. Attachment 3 provides revised (clean) TS pages. Attachment 4 provides existing TS Bases pages marked to show the proposed changes for information only.
Approval of the proposed amendment is requested by DATE to support the spring refueling outage for Unit 1. My Power & Light will implement the amendment within 90 days of the NRC approval date.
In accordance with [[CFR::10 CFR 50.91#|10 CFR 50.91,]] My Power & Light is notifying the State of [name of state] of this request by transmitting a copy of this letter and enclosure to the designated State Official.
If there are any questions or if additional information is needed, please contact Mr. I. M. Licensing at 000-111-2222 or iml@mpl.com.
I declare under penalty of perjury that the foregoing is true and correct. Executed on DATE.
I. R. Boss, Vice President
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Enclosure: Description and Assessment of the Proposed Changes
Attachment 1: List of Regulatory Commitments
Attachment 2: Existing TS Pages Marked to Show the Proposed Changes
Attachment 3: Revised (Clean) TS Pages
Attachment 4: Existing TS Bases Pages Marked to Show the Proposed Changes
cc: NRC Region X
NRC Project Manager
NRC Resident Inspector
State of [state]
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ENCLOSURE
Description and Assessment of the Proposed Changes
Subject: License Amendment Request: Add Action for Two Inoperable CREATCS Trains
1. SUMMARY DESCRIPTION
2. DETAILED DESCRIPTION
2.1 System Design and Operation
2.2 Current Technical Specifications Requirements
2.3 Reason for the Proposed Change
2.4 Description of the Proposed Change
3. TECHNICAL EVALUATION
4. REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements/Criteria
4.2 Precedent
4.3 No Significant Hazards Consideration Determination Analysis
4.4 Conclusions
5. ENVIRONMENTAL CONSIDERATION
6. REFERENCES
ATTACHMENTS:
1. List of Regulatory Commitments
2. Technical Specification Page Markups
3. Bases Page Markups (for information only)
4. Retyped Technical Specification Pages
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1 SUMMARY DESCRIPTION
The proposed change will revise [[TS::TS-3.7.11|Technical Specification (TS) 3.7.11]], "Control Room Emergency Air Temperature Control System (CREATCS)," to modify the TS Actions for two inoperable CREATCS trains. The revised Action provides 24 hours to restore a CREATCS train to operable status provided the control room area temperature is maintained below 90°F. The Action is modified by a Note, which makes the Action not applicable if the condition is entered intentionally.
2 DETAILED DESCRIPTION
2.1 System Design and Operation
The My-Plant Unit 1 and Unit 2 common control room must be kept habitable for the operators stationed there during normal operation, anticipated [[Transient|transient]]s, and design basis accidents ([[DBA|DBA]]s). The control room function is supported by two systems:
* The control room temperature is maintained by the CREATCS to support habitability for the operators and operability of the control room equipment. The CREATCS consists of two independent and redundant trains that provide cooling of recirculated control room air. A single CREATCS train can provide the required temperature control. The CREATCS is a standby system that shares components (ducting, etc.) with an air conditioning system used during normal operation. The CREATCS is not started automatically. The CREATCS is not explicitly credited in the accident analysis, but the analyses assume that the control room temperature supports control room habitability and that equipment located in the control room is maintained at a temperature that supports its operability. The CREATCS is required to be operable by [[TS::TS-3.7.11|TS 3.7.11]] in Modes 1, 2, 3 and 4. The CREATCS is also be required to be operable in Modes 5 and 6 and during movement of recently irradiated fuel assemblies.
* The [[system::Control Room Emergency Ventilation|Control Room Emergency Ventilation]] System ([[CREVS|CREVS]]) removes radioactive materials, hazardous chemicals, and smoke from the control room air. The [[CREVS|CREVS]] is a standby system that consist of two independent, redundant trains that recirculate and filter the air in the control room envelope ([[CRE|CRE]]) following an accident or [[Transient|transient]]. The system has filters, absorbers, ductwork, dampers, doors and a fan. The [[CREVS|CREVS]] is explicitly credited in the accident analysis for airborne radiological protection for the control room occupants, as well as operator protection from smoke and hazardous chemicals. The [[CREVS|CREVS]] is required to be operable by [[TS::TS-3.7.10|TS 3.7.10]] in Modes 1, 2, 3 and 4, and in Modes 5 and 6 and during movement of recently irradiated fuel assemblies. The proposed change does not alter the requirements on the [[CREVS|CREVS]].
The proposed change does not alter the TS for the [[CREVS|CREVS]]; only the TS for the CREATCS are proposed to be revised. The CREATCS and the [[CREVS|CREVS]] share components, such as ductwork, dampers, and doors. Inoperability of the CREATCS would not affect the operability of the [[CREVS|CREVS]] unless a shared component, such as ductwork, is affected.
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2.2 Current Technical Specifications Requirements
The current CREATCS TS provides an action that allows 30 days to restore an inoperable train. The 30 day Completion Time is based on the ability of the remaining operable CREATCS train to maintain the control room temperature within limits, the low probability of an event requiring control room isolation, and alternate cooling means that may be available. If the inoperable train is not restored within the 30 day Completion Time and the plant is in Modes 1, 2, 3, or 4, a plant shutdown is required. If the inoperable train is not restored within 30 days while in Modes 5 or 6, or during movement of recently irradiated fuel, movement of recently irradiated fuel assemblies must be suspended immediately.
The current CREATCS TS action for two inoperable CREATCS trains in Modes 1, 2, 3, and 4, Action E, requires entering [[TS::TS-3.0.3|LCO 3.0.3]] immediately. If two CREATCS trains are inoperable in Modes 5 or 6, or during movement of recently irradiated fuel assemblies, Action D requires movement of recently irradiated fuel assemblies to be suspended immediately.
2.3 Reason for the Proposed Change
The TS requirement to enter [[TS::TS-3.0.3|LCO 3.0.3]] and shutdown the unit immediately when two CREATCS trains are inoperable is not commensurate with the safety function provided by the system and is inconsistent with similar TS requirements.
The Current TS Action is Not Commensurate with the Safety Function
The CREATCS is not directly credited with preventing or mitigating an accident in the safety analysis. Unavailability of the CREATCS will not directly impact plant safety provided actions are in place to ensure operator habitability and equipment operability. Plant staff can monitor control room temperature to ensure it remains habitable and that electrical cabinets are not exposed to excessive temperatures. This could include actions such as use of normal (i.e., non-safety) ventilation systems, opening cabinet doors, use of fans or ice vests, and opening control room doors or ventilation paths. Therefore, requiring an immediate plant shutdown (a plant [[Transient|transient]]) is not commensurate with the level of degradation associated with two inoperable CREATCS trains.
The Requirements are not Consistent with Other Technical Specifications
There are a number My-Plant Unit 1 and Unit 2 TS that provide a 30 day Completion Time for an inoperable train and also provide an extended Completion Time for two inoperable trains. Systems are provided with a 30 day Completion Time to restore an inoperable train because the system is of low safety significance or is only relied on for low probability events.
* [[TS::TS-3.0.9|LCO 3.0.9]] permits one or more required barriers to be unable to perform their related support function(s) for up to 30 days without declaring the supported systems inoperable. Should a required redundant system be rendered inoperable, [[TS::TS-3.0.9|LCO 3.0.9]] provides 24 hours to restore it to operable status.
* [[TS::TS-3.3.17|TS 3.3.17]], "Post Accident Monitoring," provides 30 days to restore one or more functions with one required channel inoperable and 7 days to restore one or more functions with two required channels inoperable.
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* [[TS::TS-3.4.15|TS 3.4.15]], "[[RCS|RCS]] Leakage Detection Instrumentation," provides 30 days to restore one inoperable leakage monitor and either 30 days or 7 days to restore one of two inoperable leakage monitors.
All My-Plant Unit 1 and Unit 2 TS, except CREATCS, in which a 30 day Completion Time is provided for one inoperable train, time is also provided to restore one train when both trains are inoperable.
In addition to the reasons described above, situations have occurred in which both CREATCS trains were inoperable. For example:
* In 2014, My-Plant Unit 1 began a shutdown required by the TS when one train of the control room temperature control system failed while the redundant train was inoperable for maintenance. One train of control room temperature control was restored in approximately six hours, by which time the reactor power had been reduced to 33%. The proposed change would have prevented the power reduction.
* In 2011, My-Plant Unit 1 and Unit 2 entered [[TS::TS-3.0.3|LCO 3.0.3]] and shut down when one train of the control room temperature control system failed while the redundant train was inoperable for maintenance. My-Plant Units 1 and 2 share a common control room.
2.4 Description of the Proposed Change
New Conditions C and D are added:
CONDITION
REQUIRED ACTION
COMPLETION TIME
C. ----------- NOTE -----------
Not applicable when second CREATCS train intentionally made inoperable.
--------------------------------
Two CREATCS trains inoperable.
C.1 Verify control room area temperature < 90°F.
AND
C.2 Restore one CREATCS train to OPERABLE status.
Once per 4 hours
24 hours
D. Required Actions and associated Completion Times of Condition C not met in MODE 1, 2, 3, or 4.
D.1 Be in MODE 3.
AND
D.2 Be in MODE 5.
6 hours
36 hours
The existing Action C is renamed Action E.
The existing Action D is renamed Action F and revised.
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[[FD|FD]].Required Action and associated Completion Time of Condition C not met in MODE 5 or 6, or during movement of recently irradiated fuel assemblies. Two CREATCS trains inoperable during movement of recently irradiated fuel assemblies.
[[FD|FD]].1 Suspend movement of recently irradiated fuel assemblies.
Immediately
The existing Action E, which is applicable when two CREATCS trains are inoperable in Modes 1, 2, 3, or 4, is deleted.
The proposed change is supported by changes to the TS Bases. In addition to reflecting the proposed changes to the TS, the [[TS::TS-3.7.11|TS 3.7.11]] Bases are revised for clarity and consistency. The regulation at Title 10 of the Code of Federal Regulations (10 CFR), Part 50.36, states, "A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications." Changes to the TS Bases will be made in accordance with the Technical Specifications Bases Control Program following approval of the requested amendment. The proposed TS Bases changes are consistent with the proposed TS changes and provide the purpose for each requirement in the specification consistent with the Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 2, 1993 ([[FR::58 FR 39132|58 FR 39132]]). Therefore, the Bases changes are provided for information and approval of the Bases is not requested.
3 TECHNICAL EVALUATION
The proposed change provides two Required Actions to be followed when two CREATCS trains are inoperable. The first Required Action states that control room area temperature will be verified to be below 90°F once per hour. The purpose of the 90°F limit in Required Action C.1 is to support control room operator habitability and the operability of equipment in the control room. The 90°F limit in Required Action C.1 is based on maintaining control room temperature with the normal, non-safety, control room cooling system. Evaluations have determined that the 90°F limit will ensure that equipment in the control room will remain operable. The temperature limit was compared to the guidance in [[EPRI|EPRI]] [[TR|TR]] 109445, "Heat Stress Management Program for Power Plants", and it was determined that no limitations on operator stay times based on the control room temperature limit and assumed humidity levels are required.
The proposed change requires verification that the control room area temperature is less than the 90°F limit once per hour. The verification frequency of 1 hour does not allow temperature to exceed the limit between performances. As stated in the [[TS::TS-3.0.2|LCO 3.0.2]] Bases, "An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists
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or the unit is not within the LCO Applicability." There are many Required Actions in the [[ISTS|ISTS]] that verify a condition is met with Completion Times stated as "Once per...". For example, [[TS::TS-3.6.3|TS 3.6.3]], "Containment Isolation Valves," requires isolation of a penetration flow path when a containment valve in the flow path is inoperable. There are Required Actions to verify the penetration flow path is isolated once per 31 days. These Required Actions state a condition that must be maintained at all times while in the condition, not only once per 31 days.
The one-hour monitoring frequency of control room area temperature is adequate given the indications available in the control room and the time required for a significant increase in control room air temperature. When a plant is in a Required Action that can result in a plant shutdown, operator and plant management attention is on resolving the condition and satisfying the Required Actions to prevent a shutdown. Control room area temperature data is available from the control room based on four monitors distributed throughout the shared control room. My Power & Light commits to maintaining the functionality of at least three control room temperature monitors when the proposed Required Action is in use. Requiring the temperature to be recorded to document performance of the Required Action more frequently than once per hour would unnecessarily distract the operator from more safety significant activities.
The second Required Action requires at least one CREATCS train to be restored to operable status within 24 hours. Twenty-four hours is sufficient time in most circumstances to restore at least one CREATCS train to operable status while minimizing the length of time in which the CREATCS is inoperable. The small likelihood of an event requiring the CREATCS during the 24 hour Completion Time, combined with mitigating actions which can be taken to maintain control room temperature, support providing a limited time to restore an inoperable CREATCS train and potentially avoiding a [[Transient|transient]] associated with a plant shutdown.
Should a component required by the CREATCS and the [[CREVS|CREVS]] be unable to perform its required function, both [[TS::TS-3.7.10|LCO 3.7.10]] and [[TS::TS-3.7.11|LCO 3.7.11]] would be declared not met and all applicable Actions would be followed. If the inoperability affected both trains of CREATCS and both trains of [[CREVS|CREVS]], then proposed [[TS::TS-3.7.11|TS 3.7.11]] Condition C would apply, as well as [[TS::TS-3.7.10|TS 3.7.10]] Condition B, E, or F. [[TS::TS-3.7.10|TS 3.7.10]], Condition F requires immediately entering [[TS::TS-3.0.3|LCO 3.0.3]].
Should the control room area temperature not be maintained below the temperature limit or if a CREATCS train is not restored to operable status within 24 hours while in Modes 1, 2, 3, or 4, the plant must be in Mode 3 in 6 hours and Mode 5 in 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems.
Should the control room area temperature not be maintained below the temperature limit or if a CREATCS train is not restored to operable status within 24 hours while in Modes 5 or 6, or during movement of recently irradiated fuel assemblies, movement of recently irradiated fuel assemblies must be suspended immediately. The only accident considered in Modes 5 or 6 is a fuel handling accident, which does not assume a loss of offsite electrical power or additional failures, and the measures taken to maintain control room temperature within the limit will still be available should a fuel handling accident occur. The Required Action minimizes the potential for a radioactive release from a fuel handling accident which might require control room isolation and subsequent cooling.
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The Condition for two inoperable CREATCS trains is not intended to be used for planned maintenance or for intentional entry. It is modified by a Note which states, "Not applicable when second CREATCS train intentionally made inoperable." As stated in the associated Bases:
The Condition is modified by a Note stating it is not applicable when the second CREATCS train is intentionally made inoperable. This Required Action is not intended for voluntary removal of redundant systems or components from service. The Required Action is only applicable if one CREATCS train is inoperable for any reason and a second CREATCS train is found to be inoperable, or if two CREATCS trains are found to be inoperable at the same time.
Condition F, now Condition D, is revised to state, "Required Action and associated Completion Time of Condition C not met in MODE 5 or 6, or during movement of recently irradiated fuel assemblies," instead of "Two CREATCS trains inoperable during movement of recently irradiated fuel assemblies." This change provides 24 hours to restore one CREATCS train to operable status, if control room area temperature is maintained below the temperature limit, before suspending movement of recently irradiated fuel assemblies. This change is acceptable because of the small likelihood of a fuel handling accident requiring control room isolation and use of the CREATCS during the proposed 24 hour Completion Time, combined with mitigating actions which can be taken to maintain control room temperature. The analysis of a fuel handling accident does not require consideration of a loss of offsite power or an additional failure. Therefore, the mitigating actions being taken to maintain control room temperature below the limit in Required Action C.1 will continue to be available should a fuel handling accident occur. The small safety impact associated with the proposed allowance is offset by avoiding the potential safety consequences (such as the increased risk of a human performance error) that could result from the disruption of a carefully planned refueling outage schedule due to suspending movement of recently irradiated fuel assemblies.
4 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements/Criteria
The following NRC requirements and guidance documents are applicable to the proposed change.
The regulations at Title 10 of the Code of Federal Regulations (10 CFR) Part 50.36 "Technical specifications," establish the requirements related to the content of the TS. Section 50.36(c)(2) states:
Limiting conditions for operation. Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.
The regulatory requirements in [[CFR::10 CFR 50.36#|10 CFR 50.36]] are not specific regarding the actions to be followed when TS requirements are not met other than a plant shut down. The proposed change provides remedial actions in the Technical Specifications to be followed when the Limiting
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Condition for Operation is not met. Therefore, the proposed change is consistent with the requirements of [[CFR::10 CFR 50.36#|10 CFR 50.36]].
The construction permit for My-Plant Unit 1 and 2 was issued by the Atomic Energy Commission ([[AEC|AEC]]) on April 1, 1967, and an Interim Provisional Operating License was issued by the [[AEC|AEC]] on April 1, 1971. In the request for a full term operating license My Power & Light provided a discussion to compare the plant design with the General Design Criteria (GDC) as they appeared in [[CFR::10 CFR 50 Appendix A#|10 CFR 50 Appendix A]] on July 7, 1971. It was this discussion, including the identified exceptions, which formed the original plant licensing basis for compliance with the GDC. This discussion is contained in [[Licensing Basis Document::Updated Final Safety Analysis Report|Updated Final Safety Analysis Report]] ([[UFSAR|UFSAR]]) Chapter 5.1, "General Design Criteria," with more details provided in other [[UFSAR|UFSAR]] sections. As described in [[UFSAR|UFSAR]] Section 5.1.1, changes have been made to the original [[UFSAR|UFSAR]] GDC discussions to reflect commitments and changes made to the facility over the life of the plant. Therefore, the GDC discussions in the [[UFSAR|UFSAR]] constitute the My-Plant Unit 1 and 2 licensing bases with respect to compliance with the GDC. These criteria are referenced in Chapter 3 of the My-Plant Unit 1 and Unit 2 [[Licensing Basis Document::Updated Final Safety Analysis Report|Updated Final Safety Analysis Report]] ([[UFSAR|UFSAR]]). Criterion 19 states:
Criterion 19—Control room. A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 [[Rem|rem]] whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.
Applicants for and holders of construction permits and operating licenses under this part who apply on or after January 10, 1997, applicants for design approvals or certifications under part 52 of this chapter who apply on or after January 10, 1997, applicants for and holders of combined licenses or manufacturing licenses under part 52 of this chapter who do not reference a standard design approval or certification, or holders of operating licenses using an alternative source term under § 50.67, shall meet the requirements of this criterion, except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv (5 [[Rem|rem]]) total effective dose equivalent ([[TEDE|TEDE]]) as defined in § 50.2 for the duration of the accident.
The proposed change has no effect on the design of the control room or on operator radiation dose, as that protection is provided by other systems required by the Technical Specifications. The proposed change also has no effect on alternate control locations outside of the control room. Therefore, the only aspect of [[General design criterion::GDC 19|GDC 19]] applicable to the proposed change is the criterion to design the control room from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. The proposed change has no effect on the design of the
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control room and the proposed actions will ensure that the control room temperature is maintained such that the plant may be operated safely from the control room.
[[Regulatory Guide::Regulatory Guide 1.196|Regulatory Guide 1.196]], "Control Room Habitability at Light-Water Nuclear Power Reactors," Revision 1, provides guidance and criteria that the NRC staff considers acceptable for implementing the agency’s regulations as they relate to control room habitability. The Regulatory Guide addresses radiological, hazardous chemical, or smoke challenges that could result in the inability of the operators to control the reactor from the control room. It does not address the performance of the reactor controls and instrumentation systems that are affected by environmental conditions, nor does it address human engineering (i.e., temperature, vibration, sound, or lighting). Therefore, the proposed change has no effect on the application of the Regulatory Guide.
The proposed change does not affect plant compliance with these regulations or guidance and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met.
4.2 Precedent
The proposed change is consistent with NRC-approved license amendment issued to Good Power & Light on January 11, 2005 (Amendment Numbers 100/100) for Good Units 1 & 2 (NRC [[ADAMS|ADAMS]] Accession No. [[document::ML99999999|ML99999999]]). The approved TS changes are identical to the changes proposed in this request. There are no differences between the plant design and licensing basis for My-Plant Units 1 and 2 and Good Units 1 and 2 that would affect the applicability of the change.
4.3 No Significant Hazards Consideration Analysis
My Power & Light requests approval of a change to the My-Power Unit 1 and Unit 2 Technical Specifications (TS). The proposed change will [[TS::TS-3.7.11|TS 3.7.11]], "Control Room Emergency Air Temperature Control System (CREATCS)," to modify the TS Actions for two inoperable CREATCS trains. The revised Action provides 24 hours to restore a CREATCS train to operable status provided the control room area temperature is maintained below 90°F. The Action is modified by a Note which makes the Action not applicable if the condition is entered intentionally.
My Power & Light has evaluated whether a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in [[CFR::10 CFR 50.92#|10 CFR 50.92,]] "Issuance of amendment," as discussed 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 revises the CREATCS TS to modify the Actions for two inoperable CREATCS trains and provides 24 hours to restore a CREATCS train to operable status provided the control room area temperature is maintained below a limit. The CREATCS
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is not an initiator of any accident previously evaluated. As a result, the probability of an accident previously evaluated is not increased. The consequences of an accident during the proposed 24 hour Completion Time are no different than the consequences of an accident in Modes 1, 2, 3, and 4 during the existing 1 hour Completion Time provided in [[TS::TS-3.0.3|LCO 3.0.3]] to prepare for a shutdown. The only accident previously evaluated in Modes 5 or 6 is a fuel handling accident. The accident evaluation does not assume a loss of offsite electrical power or additional failures, and the measures taken to maintain control room temperature within the limit will still be available should a fuel handling accident occur. As a result, providing 24 hours to restore one train of control room cooling does not significantly increase the consequences of a fuel handling accident over the current requirement.
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 previously evaluated?
Response: No
The proposed change revises the CREATCS TS to modify the Actions for two inoperable CREATCS trains and provides 24 hours to restore a CREATCS train to operable status provided the control room area temperature is maintained below a limit. The proposed change will not alter the design or function of the control room or the CREATCS. Should the new Actions not be met, the existing and proposed Actions require a plant shutdown.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No
The proposed change revises the CREATCS TS to modify the Actions for two inoperable CREATCS trains and provides 24 hours to restore a CREATCS train to operable status provided the control room area temperature is maintained below a limit. The change provides a limited period of time to restore an inoperable CREATCS train instead of requiring an immediate plant shutdown or suspension of movement of irradiated fuel assemblies. A plant shutdown is a [[Transient|transient]] that may be avoided by providing a limited time to make repairs. In addition, the control room area temperature must be maintained less than a limit set to ensure habitability of the control room and the operability of the equipment cooled by the CREATCS. The potential to avoid a plant [[Transient|transient]] in conjunction with maintaining the control room temperature offsets any risk associated with the limited Completion Time. Any small impact on the safety margin associated with immediately suspending movement of irradiated fuel assemblies is offset by avoiding the potential safety consequences (such as the increased risk of a human
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performance error) that could result from the disruption of a carefully planned refueling outage schedule due to suspending fuel movement.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, My Power & Light concludes that the proposed change presents no significant hazards consideration under the standards set forth in [[CFR::10 CFR 50.92#c|10 CFR 50.92(c)]], and, accordingly, a finding of "no significant hazards consideration" is justified.
4.4 Conclusion
In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission’s regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
5 ENVIRONMENTAL CONSIDERATION
A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in [[CFR::10 CFR 20#|10 CFR 20,]] or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in [[CFR::10 CFR 51.22#c9|10 CFR 51.22(c)(9)]]. Therefore, pursuant to [[CFR::10 CFR 51.22#b|10 CFR 51.22(b)]], no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
6 REFERENCES
1. U.S. Code of Federal Regulations, Appendix A, "General Design Criteria."
2. U.S. NRC to Good Power & Light, "License Amendment Approval," dated January 11, 2005, NRC [[ADAMS|ADAMS]] Accession No. [[document::ML99999999|ML99999999]].
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Attachment 1
List of Regulatory Commitments
The following table identifies the regulatory commitments in this document. Any other statements in this submittal represent intended or planned actions, are provided for information purposes, and are not considered to be regulatory commitments.
COMMITMENT
TYPE
SCHEDULED COMPLETION DATE
(if applicable)
One-time
Continuing Compliance
My Power & Light commits to maintaining the functionality of at least three control room area temperature monitors when the proposed Required Action C.1 is in use.
X
On implementation of the approved amendment.
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Attachment 2
Existing TS Pages Marked to Show the Proposed Changes
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Attachment 3
Revised (Clean) TS Pages
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Attachment 4
Existing TS Bases Pages Marked to Show the Proposed Changes
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C-1

=APPENDIX C EXIGENT/EMERGENCY LARS=
BACKGROUND
When a licensee submits a [[LAR|LAR]], the NRC is required by [[CFR::10 CFR 50.91#|10 CFR 50.91]] to publish a "notice of opportunity for hearing" in the Federal Register at least 30 days before it issues the amendment. However, circumstances can occur where an amendment is warranted to preclude an unnecessary plant [[Transient|transient]], test, inspection, system realignment, shutdown, or delayed startup. If this is the case, a licensee may request an "exigent" amendment or an "emergency" amendment.
Exigent circumstances exist if the NRC must act quickly and time does not permit the normal 30-day public comment period. Emergency circumstances exist if NRC must act immediately and public comment must be postponed until after the amendment is issued. Licensees should be prepared to explain the need for expedited approval, why it could not be avoided, and why the need for a routine amendment could not be foreseen in advance.
The exigent amendment is the preferred alternative because it provides for a reduced (typically 14-day) public comment period prior to issuance of the amendment. If issuance is necessary in less than 14 days, the NRC may publish a notice in local media in the area of the plant.
Typically, the need for an exigent or emergency amendment occurs when an unforeseen situation prevents the licensee from satisfying a TS LCO and the plant must be shut down if conformance with the LCO cannot be reestablished within a specified time limit (the Completion Time). The process for either alternative is time limited and highly variable. A request for an exigent or emergency amendment is feasible if supporting information is available and the proposed amendment can be prepared and approved by the NRC before the Completion Time expires. The longer it takes to compile and submit the necessary information, the lower the probability of obtaining NRC approval in time.
REGULATORY CRITERIA
[[CFR::10 CFR 50.91#a5|10 CFR 50.91(a)(5)]]
"Where the Commission finds that an emergency situation exists, in that failure to act in a timely way would result in derating or shutdown of a nuclear power plant, or in prevention of either resumption of operation or of increase in power output up to the plant’s licensed power level, it may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for a hearing or for public comment. In such a situation, the Commission will not publish a notice of proposed determination on no significant hazards consideration, but will publish a notice of issuance under § 2.106 of this chapter, providing for opportunity for a hearing and for public comment after issuance. The Commission expects its licensees to apply for license amendments in timely fashion. It will decline to dispense with notice and comment on the determination of no significant hazards consideration if it determines that the licensee has abused the emergency provision by failing to make timely application for the amendment and thus itself creating the emergency. Whenever an emergency situation exists, a licensee requesting
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an amendment must explain why this emergency situation occurred and why it could not avoid this situation, and the Commission will assess the licensee’s reasons for failing to file an application sufficiently in advance of that event."
[[CFR::10 CFR 50.91#a6|10 CFR 50.91(a)(6)]]
"Where the Commission finds that exigent circumstances exist, in that a licensee and the Commission must act quickly and that time does not permit the Commission to publish a Federal Register notice allowing 30 days for prior public comment, and it also determines that the amendment involves no significant hazards considerations, it:
(i)(A) Will either issue a Federal Register notice providing notice of an opportunity for hearing and allowing at least two weeks from the date of the notice for prior public comment; or
(B) Will use local media to provide reasonable notice to the public in the area surrounding a licensee’s facility of the licensee’s amendment and of its proposed determination as described in paragraph (a)(2) of this section, consulting with the licensee on the proposed media release and on the geographical area of its coverage;
(ii) Will provide for a reasonable opportunity for the public to comment, using its best efforts to make available to the public whatever means of communication it can for the public to respond quickly, and, in the case of telephone comments, have these comments recorded or transcribed, as necessary and appropriate;
(iii) When it has issued a local media release, may inform the licensee of the public’s comments, as necessary and appropriate;
(iv) Will publish a notice of issuance under § 2.106;
(v) Will provide a hearing after issuance, if one has been requested by a person who satisfies the provisions for intervention specified in § 2.309 of this chapter;
(vi) Will require the licensee to explain the exigency and why the licensee cannot avoid it, and use its normal public notice and comment procedures in paragraph (a)(2) of this section if it determines that the licensee has failed to use its best efforts to make a timely application for the amendment in order to create the exigency and to take advantage of this procedure."
GUIDANCE TO LICENSEES
Overview
Once the decision is made to pursue an emergency or exigent TS Amendment, a call should be conducted with the NRC to alert them to the situation and to confirm that the emergency or exigent [[LAR|LAR]] process is the appropriate regulatory process. Typically, this call is initiated by the licensee contacting the NRC [[PM|PM]], though a courtesy notification of the NRC resident staff is also prudent.
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Early and frequent communication with the NRC (NRC [[PM|PM]] and NRC Senior Resident Inspector) is important to completion of the process due to the time constraints involved. Early communication allows the NRC [[PM|PM]] to align the appropriate resources in [[NRR|NRR]] to support review of the proposed amendment and discuss with the licensee schedule and contacts. The licensee should inform the NRC of key information, such as the applicable Technical Specification or design bases changes, descriptions/drawings of the associated systems, a brief timeline of the event, discussion of the apparent cause, activities underway, contingencies in place should the amendment not be approved, etc. Any security considerations should also be included, and whether the changes affect the emergency or security plans, or the fire protection plan.
Preparation and approval of an emergency TS [[LAR|LAR]] requires many critical path activities which are performed by several different groups (see Table C.1). For that reason, teamwork and open and timely communication must be maintained throughout the preparation and approval process.
Timing
The guidance in Table C.1 is based on a "best estimate" of the most likely time to complete each step in the process of preparing an exigent or emergency [[LAR|LAR]]. If the steps cannot be completed in accordance with this schedule, the licensee should revisit the decision to pursue the exigent or emergency amendment.
In addition, the following factors affect the feasibility of, or need for, an exigent/emergency LAR:
* The time available to complete the [[LAR|LAR]] before the Completion Time expires
* The likelihood of meeting the LCO before the Completion Time expires
* The availability of [[PRA|PRA]] calculations that provide an appropriate risk justification
o Note: the NRC will assess the quality of [[PRA|PRA]] against [[Regulatory Guide::Regulatory Guide 1.200|Regulatory Guide 1.200]] (Ref. 13)
* The likelihood and timing of approvals by the onsite and offsite review boards
* The feasibility of a [[LAR|LAR]] as an alternative to a [[Notice of Enforcement Discretion|Notice of Enforcement Discretion]] ([[NOED|NOED]]), though [[LAR|LAR]]s are viewed by the NRC as preferable whenever possible
* The status of shutdown preparations
* NRC feedback.
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TABLE C-1 Estimated Time to Prepare an Exigent or Emergency License Amendment Request
Activity
Description
Estimated Time (hours)
A
Assess the likelihood that the TS LCO will be met before the Completion Time expires. This activity is highly variable and should receive increased management attention.
8
B
Select and initiate the appropriate regulatory process.
1/3
C
Contact [[PRA|PRA]] specialists and prepare documentation
2
D
Complete [[PRA|PRA]] documentation in support of [[PRA|PRA]] calculations (e.g., equipment out of service at the time of the event, available station equipment, a detailed time line for the event, event causal factors, and extent of condition). This activity is highly variable and should receive management attention.
8
E
Prepare [[PRA|PRA]] calculations. This activity requires site-specific information and a checked/approved calculation.
10
F
Incorporate [[PRA|PRA]] data into the [[PRA|PRA]] calculation.
4
G
Incorporate [[PRA|PRA]] calculation into the [[LAR|LAR]].
4
H
Prepare the [[LAR|LAR]] package, including technical review and concurrence.
12
I
Obtain review by the onsite and offsite review committees as required by plant programs and procedures.
2
J
Obtain NRC review of the [[LAR|LAR]].
12
(Minimum)
Exigent/Emergency [[LAR|LAR]] Template
The following information specific to exigent or emergency amendments should be used in addition to the guidance given in Appendix A.
COVER LETTER
"Approval of the proposed amendment is requested by [date + justification] and will remain in effect for [time period]."[Application should clearly denote whether the requested change is "one
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time only" or is a permanent change to the OL/TS. "One time only" requests should justify the duration.]
"Attachment 2 is a placeholder for the reprinted TS pages reflecting the proposed changes. The reprinted TS changes will be provided to the NRC Project Manager pending the completion of NRC review of this request."
"In accordance with the licensee administrative procedures and the Quality Assurance Program Manual, this proposed amendment has been previously reviewed and approved by the Plant Operations Review Committee and the Nuclear Safety Review Board."
DETAILED DESCRIPTION
Discuss the conditions that the proposed amendment is intended to resolve.
Explain the circumstances that establish a need for the proposed amendment, including why the situation occurred and why it could not be avoided (as required by [[CFR::10 CFR 50.91#a5|10 CFR 50.91(a)(5)]]).
TECHNICAL EVALUATION
Discuss the following, if appropriate, for the requested change:
* What [[SSC|SSC]]s will be protected during the extended period?
* What can affect grid stability and what actions have been taken to preserve grid stability?
* What is the cause of failure?
* What is the timeline for the repair?
* Evidence that the repair plan will be successful.
* Clarification that no activities either in progress or planned will affect any of the conditions assumed by [[PRA|PRA]] to justify their conclusions. This is related to protected equipment which preserves the risk assumed for the duration of the extension.
* The effect on [[UFSAR|UFSAR]] accident analysis
* A discussion of the technical aspects of relevant precedents, including reference to the discussion of precedent in 4.2
Note: Consider adding compensatory measures to support the request.
It is important to note that the amendment must stand on its own merit; despite the collapsed review time, and all the same standards apply for the NRC safety review.
NOTE: Determine the critical path activities. All activities must be completed before the limiting TS Completion Time expires. Activity descriptions are high-level and may involve
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multiple tasks. For example, Activity B involves contacting NRC and onsite/offsite review boards to alert them of the situation.
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=APPENDIX D PLANT-SPECIFIC ADOPTION OF [[TSTF|TSTF]] TRAVELERS=
INTRODUCTION
The [[TSTF|TSTF]], a jointly sponsored activity of the [[PWR|PWR]] and [[BWR|BWR]] Owners Groups, and the AP1000® Owners Group, develops generic changes to the Improved Standard Technical Specifications ([[ISTS|ISTS]]). The changes are called "travelers." Travelers are submitted to the NRC for review. After a traveler is approved by NRC, it is given an "A" postscript (e.g., TSTF-445-A) and posted on the [[TSTF|TSTF]] web site (http://www.excelservices.com).
Most travelers approved by the NRC since 2000 have a model application to be used by utilities adopting the change and a model safety evaluation to facilitate documentation of the NRC staff review. [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-600|LIC-600]], "Review of Technical Specifications Task Force ([[TSTF|TSTF]]) Travelers and Creation of '[[CLIIP|CLIIP]]' Model Applications," (Reference 14) describes the overall NRC process for review and approval of travelers. This includes the budgeting and scheduling of NRC resources for traveler reviews, the coordination of NRC technical staff review and concurrence, and the drafting of a Model [[SE|SE]] for each traveler.
The NRC's approval of some [[TSTF|TSTF]] travelers state they are available under the Consolidated Line Item Improvement Process ([[CLIIP|CLIIP]]). License amendments to adopt [[CLIIP|CLIIP]] travelers are straightforward and do not require review by the NRC's technical branches. They are reviewed only by the Technical Specifications Branch and the NRC's internal goal is to complete review of the [[CLIIP|CLIIP]] [[LAR|LAR]]s within 6 months. If the NRC reviewer identifies a significant change from the approved traveler, the [[LAR|LAR]] is removed from the [[CLIIP|CLIIP]] process and distributed for review by the NRC technical branches, similar to a non-CLIIP change.
[[NRR|NRR]] Office Instruction [[NRC LIC::LIC-101|LIC-101]] describes the overall NRC process for managing [[LAR|LAR]] reviews, including [[LAR|LAR]]s based on travelers. The adoption of [[TSTF|TSTF]] travelers promotes consistency among plant-specific TS. There are hundreds of NRC-approved changes to the [[ISTS|ISTS]], most of which have been adopted by individual licensees by means of plant-specific [[LAR|LAR]]s or as part of conversion to the [[ISTS|ISTS]].
The options for plant-specific adoption of travelers are:
* Adoption of a single traveler using the model application.
* Adoption of a single traveler that does not have a model application.
* Adoption of a template ("T") traveler
ADOPTION OF A SINGLE TRAVELER USING THE MODEL APPLICATION
The adoption of travelers with a model application is described in this document as a Type 2 or Type 3 [[LAR|LAR]].
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When adopting a traveler with an NRC-accepted model application, the [[LAR|LAR]] should follow the NRC-approved model application as closely as the plant design and licensing basis allows. Changes not described in the "Variations" section of the model application should not be included. Changes not described in the NRC-approved model Safety Evaluation should not be included. Simple administrative (i.e., typographical corrections) or conforming changes needed to address differences between the plant-specific TS and the traveler markups are acceptable variations from the traveler. If in doubt, discuss the changes with the NRC [[PM|PM]]. If technically significant deviations from the generically approved change are needed, a Type 1 [[LAR|LAR]] should be written that does not claim to adopt the traveler (although the traveler may be cited as precedent for the applicable portions of the [[LAR|LAR]]).
The NRC provided the following examples of changes that were outside the scope of the traveler:
1. For adoption of [[TSTF::TSTF-523|TSTF-523]], "[[generic letter::NRC Generic Letter 08-01|Generic Letter 2008-01]] Managing Gas Accumulation," changes to surveillance intervals were also proposed.
2. For adoption of [[TSTF::TSTF-545|TSTF-545]], "TS In-service Testing ([[IST|IST]]) Program Removal," changes to an [[ISI|ISI]] program were also proposed.
3. For plants with custom TS, 3 /4 format TS, or unique design features, a lack of technical justification for certain changes would adversely impact the review schedule.
ADOPTION OF A SINGLE TRAVELER THAT DOES NOT HAVE A MODEL APPLICATION AND [[SE|SE]]
[[LAR|LAR]]s to adopt [[TSTF|TSTF]] travelers that do not have a model application should be requested as a Type 1 [[LAR|LAR]].
* Minimize the differences between the [[LAR|LAR]] and the NRC-approved traveler. If there are differences, they must be fully explained and justified to facilitate NRC review.
* The NRC typically did not prepare an [[SE|SE]] for approved travelers numbered less than 400. For most of these travelers, the NRC provided a letter stating the traveler was approved, but some of them were approved during public meetings without a letter being written. [[LAR|LAR]]s that reference [[TSTF|TSTF]]s below [[TSTF::TSTF-400|TSTF-400]] should provide the NRC approval date and, if available, an example of a representative plant that has adopted the traveler, including the approval date and amendment number. The [[LAR|LAR]] should also discuss any significant differences from the referenced plant-specific [[LAR|LAR]].
* In general, it is preferable to summarize the justification for an NRC-approved traveler in the [[LAR|LAR]] instead of completely restating it. The exceptions are the older travelers for which NRC approval documentation is limited and for which there may not be any adoption precedent. The first adoption of such a traveler should provide a justification for the change that supplements and is consistent with the justification provided in the traveler.
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ADOPTION OF TEMPLATE ("T") TRAVELERS
Some travelers, called T-travelers, only modify the TS Bases and do not require NRC review or were determined to be insufficiently cost-beneficial to justify Owners Group funding of NRC review fees and were not submitted to the NRC for review and approval. However, the travelers were sufficiently cost-beneficial to develop and post to the [[TSTF|TSTF]] web site for use as templates for plant-specific use. The "T" stands for "template," e.g., TSTF-445-T. The industry traveler review process ensures that T-travelers meet the same [[ISTS|ISTS]] format and usage rules as travelers that are submitted for generic approval by NRC.
T-travelers are adopted using the TS Bases Control Program or the Type 1 [[LAR|LAR]] process.
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=APPENDIX E LICENSE AMENDMENT REQUESTS WITH RISK-INFORMED JUSTIFICATION=
INTRODUCTION
All [[LAR|LAR]]s must address the deterministic aspects of a license amendment, such as the effect of the proposed change on the design and licensing basis for the plant. In addition, [[LAR|LAR]]s that use a risk-informed justification must provide the information described in described in [[Regulatory Guide::Regulatory Guide 1.174|RG 1.174]], "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." In addition, [[Regulatory Guide::Regulatory Guide 1.200|RG 1.200]], "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," provides technical standards for the [[PRA|PRA]] model used to support development of risk-informed [[LAR|LAR]]s. [[Regulatory Issue Summary::RIS 2007-06|RIS 2007-06]], "[[Regulatory Guide::Regulatory Guide 1.200|Regulatory Guide 1.200]] Implementation," ([[ADAMS|ADAMS]] Reference [[document::ML070650428|ML070650428]]), states that the NRC will use [[Regulatory Guide::Regulatory Guide 1.200|Regulatory Guide 1.200]] to assess technical adequacy of all risk-informed [[LAR|LAR]]s submitted after December 2007.
RECOMMENDED CONTENT
The basic elements of a [[LAR|LAR]], as discussed in Appendix A of this document, apply to all Type 1 and Type 2 [[LAR|LAR]]s, including those that are supported with a risk-informed justification. In addition, all risk-informed Type 1 and 2 [[LAR|LAR]]s should include the following additional elements in the technical evaluation:
1) [[Regulatory Guide::Regulatory Guide 1.174|RG 1.174]], "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," provides additional guidance. The content and level of detail of the [[LAR|LAR]] must be sufficient to support the development of the key elements of the NRC's SE:
a. Technical adequacy of the baseline [[PRA|PRA]] model and the cause/effect relation supporting the [[LAR|LAR]] meets the published standards in [[Regulatory Guide::Regulatory Guide 1.200|RG 1.200]];
i. Include description of how Peer Review Findings of the base model have been closed using an NRC endorsed closure process
ii. For all open Peer Review Findings, provide a disposition of the findings for the specific application, including any appropriate sensitivity studies.
b. All applicable risk sources are addressed (e.g., fire, seismic, other external, shutdown);
i. For risk sources for which there is not a [[Regulatory Guide::Regulatory Guide 1.200|RG 1.200]] [[PRA|PRA]] model, provide either a bounding analysis or qualitative evaluations to assess the significance of the hazard to the application.
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c. Risk metrics in [[Regulatory Guide::Regulatory Guide 1.174|RG 1.174]] are satisfied:
i. Note that the risk metrics should not be considered absolute limits. [[NUREG::NUREG-1855|NUREG-1855]], "Guidance on the Treatment of Uncertainties Associated With [[PRA|PRA]]s in Risk-Informed Decision Making," should be used to evaluate impact of uncertainties and assumptions on the risk metrics.
ii. The change in Core Damage Frequency (ΔCDF) and change in Large Early Release Frequency (ΔLERF) are small;
iii. The total mean [[CDF|CDF]] and [[LERF|LERF]] (if required) from all hazards are below the Commission’s Subsidiary Safety Goal ([[CDF|CDF]] ≤ [[SSC::1E-4|1E-4]]/year and [[LERF|LERF]] ≤ [[SSC::1E-5|1E-5]]/year);
a. Note that the change is risk is considered very small, a calculation of total [[CDF|CDF]] and [[LERF|LERF]] is not required.
d. If the risk metrics approach the acceptance criteria, the submittal should address the following issues:
i. The cumulative impact of previous changes and the trend in [[CDF|CDF]] (the risk management approach),
ii. The cumulative impact of previous changes and the trend in [[LERF|LERF]] (the risk management approach),
iii. The impact of the proposed change on operational complexity, burden on the operating staff, and overall safety practices,
iv. Plant-specific performance and other factors (for example, siting factors, inspection findings, performance indicators, and operational events), and Level 3 [[PRA|PRA]] information, if available,
v. The benefit of the change in relation to its CDF/LERF increase,
vi. The practicality of accomplishing the change with a smaller CDF/LERF impact, and
vii. The practicality of reducing CDF/LERF when there is reason to believe that the baseline CDF/LERF are above the guideline values (i.e., 10-4 and 10-5 per reactor year, respectively).
e. Significant sources of uncertainty should be identified and their impact addressed using the concepts in [[NUREG::NUREG-1855|NUREG-1855]].
f. Truncation and common cause effects are addressed;
g. Any exemptions from existing regulations should be explicitly identified;
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h. The impact of the change on safety margins and defense-in-depth should be discussed; and
i. Implementation monitoring and feedback are identified.
2) [[Regulatory Guide::Regulatory Guide 1.200|RG 1.200]] provides additional submittal documentation guidance in Section 4.2. The following recommendations are provided regarding application of [[Regulatory Guide::Regulatory Guide 1.200|RG 1.200]] to a risk-informed LAR:
a. Identify the parts of the [[PRA|PRA]] used to support the [[LAR|LAR]] and the applicable requirements from the [[PRA|PRA]] standards referenced in the RG.
b. Verify the requirements from the standards are met or disposition any requirements applicable to the [[LAR|LAR]] that are not met, such as open Peer Review findings (F&Os), or capability category assessment deficiencies.
a. Note that if an NRC endorsed determination of applicable supporting requirements (SRs) is available (e.g., Risk Informed In-Service Testing), only open Peer Review Findings related to the relevant SRs need to be addressed.
c. Assess relevant [[PRA|PRA]] assumptions/approximations using sensitivity studies.
d. Submit documentation as described in [[Regulatory Guide::Regulatory Guide 1.200|RG 1.200]], Section 4.2, including:
i. Description and disposition of plant changes not incorporated in the [[PRA|PRA]] model
ii. A summary of the risk assessment methodology that was used
iii. Description of relevant key assumptions and approximations
iv. Identification that closed peer review/self-assessment findings were closed in accordance with an NRC accepted process or provide sufficient information to allow the NRC to close the findings.
v. Discussion of disposition of the impact of relevant peer review/self-assessment findings to the application.
3) For [[LAR|LAR]]s that request a risk-informed change to the In-Service Testing Program, the guidance in [[Regulatory Guide::Regulatory Guide 1.175|RG 1.175]], "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Inservice Testing," should be used:
a. A request to implement a risk-informed In-Service Testing ([[RI-IST|RI-IST]]) program as an authorized alternative to the current NRC-endorsed [[ASME|ASME]] Code pursuant to [[CFR::10 CFR 50.55a#a3i|10 CFR 50.55a(a)(3)(i)]].
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b. A description of the change associated with the proposed [[RI-IST|RI-IST]] program
c. Identification of the aspects of the plant's design,
i. Operations, and other activities that require NRC approval that would be changed by the proposed [[RI-IST|RI-IST]] program. This will provide a basis from which the staff can evaluate the proposed changes.
ii. Identification of the specific revisions to existing testing schedules and methods that would result from implementation of the proposed program.
iii. Identification of the components in the plant that are directly and indirectly involved with the proposed testing changes. Any components that are not presently covered in the plant's [[IST|IST]] program but are determined to be important to safety (e.g., through [[PRA|PRA]] insights) should also be identified. The systems that are affected by the proposed changes should be identified since this information is an aid in planning the supporting engineering analyses.
iv. Identification of the information that will be used in support of the changes. This will include performance data, traditional engineering analyses, and [[PRA|PRA]] information.
v. A brief statement describing the way how the proposed changes meet the objectives of the Commission's [[PRA|PRA]] Policy Statement.
vi. A description of the change associated with the proposed [[RI-IST|RI-IST]] program
vii. Identification of any changes to the plant's design, operations, and other activities associated with the proposed [[RI-IST|RI-IST]] program and the basis for the acceptability of these changes.
viii. A summary of key technical and administrative aspects of the overall [[RI-IST|RI-IST]] program that includes:
a. A description of the process used to identify candidates for reduced and enhanced [[IST|IST]] requirements, including a description of the categorization of components using the [[PRA|PRA]] and the associated sensitivity studies;
b. A description of the [[PRA|PRA]] used for the categorization process and for the determination of risk impact, in terms of the process to ensure quality and the scope of the [[PRA|PRA]], and how limitations in quality, scope, and level of detail are compensated for in the integrated decision making process; and
c. A description of how the impact of the change is modeled in the [[IST|IST]] components (including a quantitative or qualitative treatment
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of component degradation) and a description the impact of the change on plant risk in terms of [[CDF|CDF]] and [[LERF|LERF]] and how this impact compares with the decision guidelines.
ix. A description of how the key principles were (and will continue to be) maintained
a. A description of the integrated decision making process used to help define the [[RI-IST|RI-IST]] program, including any decision criteria used.
x. A general implementation approach or plan
a. A description of the testing and monitoring proposed for each component group;
b. A description of the [[RI-IST|RI-IST]] corrective action plan; and
c. A description of the [[RI-IST|RI-IST]] program periodic reassessment plan.
xi. A summary of any previously approved relief requests for components categorized as [[HSSC|HSSC]] along with any exemption requests, technical specification changes, and relief requests needed to implement the proposed [[RI-IST|RI-IST]] Program
xii. An assessment of the appropriateness of previously approved relief requests
4) For [[LAR|LAR]]s that request a risk-informed Graded Quality Assurance (GQA) Program, the guidance in [[Regulatory Guide::Regulatory Guide 1.176|RG 1.176]], "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Graded Quality Assurance", should be followed and the existing [[QA|QA]] program description contained in, or referenced by, the [[FSAR|FSAR]] should be revised to describe the GQA program provisions. The submittal containing the proposed GQA provisions should contain the following
a. A discussion of the essential implementation elements of the GQA program, the scope of potential [[SSC|SSC]]s that may be in the GQA program, and the basis for concluding that the overall GQA program provides reasonable confidence that [[SSC|SSC]]s remain capable of performing their intended function.
b. An overview discussion of the process and guidelines developed by the licensee to determine the safety-significance categorization of all [[SSC|SSC]]s within the GQA program scope as defined in this regulatory guide.
c. A statement of the role of the staff who perform the integrated assessment function (expert panel).
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d. The process for determining the [[QA|QA]] controls being applied to each safety-significance category of [[SSC|SSC]]s.
e. A description of the adjustments proposed as part of the GQA program and how the requirements of each of the criterion of Appendix B to [[CFR::10 CFR 50#|10 CFR Part 50]] will be satisfied in a graded manner. The description should identify any exceptions to existing [[QA|QA]] program commitments (such as regulatory guides).
f. A discussion of how augmented [[QA|QA]] controls for non-safety-related [[SSC|SSC]]s categorized as high safety significant will be determined.
g. A discussion of the operational feedback and enhanced corrective action mechanisms and processes to adjust both safety-significance categorization of [[SSC|SSC]]s and the associated [[QA|QA]] controls.
h. A discussion of the performance monitoring process, along with the [[SSC|SSC]] functional performance and availability attributes that form the basis of the proposed change.
5) For [[LAR|LAR]]s that request a change to Technical Specifications Completion Times or Surveillance Frequencies, [[Regulatory Guide::Regulatory Guide 1.177|RG 1.177]], "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," provides the following additional requirements:
a. The proposed change is consistent with the defense-in-depth philosophy.
b. The proposed change maintains sufficient safety margins.
c. When proposed changes result in an increase in core damage frequency or risk, the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement, as discussed in [[Regulatory Guide::Regulatory Guide 1.174|RG 1.174]] below.
d. The impact of the proposed change should be monitored using performance measurement strategies.
e. Changes made to the [[PRA|PRA]] for use in the TS change evaluation.
f. Discussion of the risk measures used in evaluating the changes.
g. Data developed and used in addition to the plant’s [[PRA|PRA]] database.
h. Summary of the risk measures calculated including intermediate results.
i. For submittals related to TS Completion Time Changes:
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1. For permanent changes:
a. Demonstrate that the TS [[CT|CT]] change has only a small quantitative impact on plant risk. An Incremental Conditional Core Damage Probability ([[ICCDP|ICCDP]]) of less than 1.0x10-6 and an Incremental Conditional Large Early Release Probability ([[ICLERP|ICLERP]]) of less than 1.0x10-7 are considered small for a single TS condition entry.
b. Demonstrate that there are appropriate restrictions on dominant risk-significant configurations associated with the change.
c. Identify the risk-informed plant configuration control program to utilize, maintain, and control risk with equipment inoperable (typically this the program used for compliance with 10 [[CDF|CDF]] 50.65 (a)(4)).
2. For one-time changes, the same requirements apply as for permanent changes, with the exception of the [[ICCDP|ICCDP]] and [[ICLERP|ICLERP]]. These requirements are:
1. [[ICCDP|ICCDP]] of less than 1.0 x 10-6 and an [[ICLERP|ICLERP]] of less than 1.0 x 10-7, or
2. [[ICCDP|ICCDP]] of less than 1.0x10-5 and an [[ICLERP|ICLERP]] of less than 1.0x10-6 with effective compensatory measures implemented to reduce the sources of increased risk.
6) For [[LAR|LAR]]s that request a change to risk informed In-Service Inspection Programs (RI-ISI), [[Regulatory Guide::Regulatory Guide 1.178|RG 1.178]], "An Approach for Plant-Specific Risk-Informed Decisionmaking for Inservice Inspection of Piping," provides the following additional requirements:
a. A request to implement a RI-ISI program as an authorized alternative to the current NRC endorsed [[ASME|ASME]] Code pursuant to [[CFR::10 CFR 50.55a#a3i|10 CFR 50.55a(a)(3)(i)]]. The licensee should also provide a description of how the proposed change impacts any commitments made to the NRC.
b. Identification of the aspects of the plant’s current requirements that would be affected by the proposed RI-ISI program. This identification should include all commitments and augmented programs (for example, the intergranular stress corrosion cracking ([[IGSCC|IGSCC]]) inspections and other commitments arising from generic letters affecting piping integrity) that the licensee intends to change or terminate as part of the RI-ISI program. The application of the RI-ISI methodology to incorporate and change the augmented program should be justified.
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c. Identification of the specific revisions to existing inspection schedules, locations, and methods that would result from implementation of the proposed program.
d. Plant procedures or documentation containing the guidelines for all phases of evaluating and implementing a change in the [[ISI|ISI]] program based on probabilistic and traditional insights. These should include a description of the integrated decision making process and criteria used for categorizing the safety significance of piping segments, a description of how the integrated decision making was performed, a description and justification of the number of elements to be inspected in a piping segment, the qualifications of the individuals making the decisions, and the guidelines for making those decisions.
e. The results of the licensee’s ISI-specific analyses used to support the program change with enough detail to be clearly understandable to the reviewers of the program. These results should include the following information.
i. A list of the piping systems reviewed.
ii. A list of each segment, including the number of welds, weld type, and properties of the welding material and base metal, the failure potential, [[CDF|CDF]], CCDF/CCDP, [[LERF|LERF]], CLERF, importance measure results (risk achievement worth ([[RAW|RAW]]), Fussel-Vesely (F-V), etc.) and justification of the associated threshold values, degradation mechanism, test and inspection intervals used in or in support of the [[PRA|PRA]], etc. Results from other methods used to develop the consequences and categorization of each segment (or weld) should be documented in a similar level of detail.
iii. The degradation mechanisms for each segment (if segments contain welds exposed to different degradation mechanism, for each weld) used to develop the failure potential of each segment. For the selected limiting locations, provide examples of the failure mode, failure potential, failure mechanism, weld type, weld location, and properties of the welding material and base metal.
iv. A detailed description and justification for the number of elements to be inspected.
v. Equipment assumed to fail as a direct or indirect consequence of each segment’s failure (if segments contain welds with different failure consequences, for each weld).
vi. A description of how the impact of the change between the current Section XI and the proposed RI-ISI programs is evaluated or bounded, and how this impact compares with the risk guidelines in Section 2.2.2.2 of [[Regulatory Guide::Regulatory Guide 1.174|Regulatory Guide 1.174]].
f. The means by which failure probabilities, frequencies, or potential were determined.
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g. If the submittal includes modified inspection intervals, the methodology and results of the analysis should be submitted.
h. A description of the implementation, performance monitoring, and corrective action strategies and programs in sufficient detail for the staff to understand the new [[ISI|ISI]] program and its implications.
i. Reference to NRC-approved topical reports on implementing a RI-ISI and supporting documents. Variations from the topical reports and supporting documents should be clearly identified.
7) For [[LAR|LAR]]s that request a change to implement the provisions of [[CFR::10 CFR 50.69#|10 CFR 50.69,]] the NEI guidance and template provided in [[NEI::NEI 16-09|NEI 16-09]], "Risk Informed Engineering Programs ([[CFR::10 CFR 50.69#|10 CFR 50.69)]] Implementation Guidance," should be used.
Information regarding the technical adequacy of other [[PRA|PRA]] models used based on their conformance to the high level technical characteristics or attributes in [[Regulatory Guide::Regulatory Guide 1.200|RG 1.200]], Table 3.
AVOID COMMON DEFICIENCIES
The NRC has identified the following common deficiencies in risk-informed submittals and care should be taken to ensure these areas are adequately addressed:
1) Missing or incomplete description of how cause/effect was modeled.
a. Completion Time extension where [[PRA|PRA]] assumed repair of out-of-service component;
b. The use of "zero-maintenance" assumption of other equipment is not stated;
c. For systems/components not modeled in the [[PRA|PRA]], "negligible risk" is stated but not supported; and,
d. Submittals should provide detailed information on Structure, System, or Component ([[SSC|SSC]]) modeling, data, assumptions, plant-specific performance, etc., to justify the model is adequate.
2) No discussion of common cause impacts (for Technical Specification changes applicable to unplanned emergent failures).
a. [[Regulatory Guide::Regulatory Guide 1.177|RG 1.177]], Section 2.3.3, and Appendix A are not addressed in the risk calculations.
b. No identification or justification of the truncation level used.
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3) The submittal should state the assumption, provide a basis for why it is valid and reasonable, and explore potential risk impacts with sensitivity studies. Stating an assumption of believed to be reasonable is insufficient.
4) Justification of open findings and observations inadequate.
a. Unsupported statements are made, such as "judged not to impact the risk calculations," "no impact" is insufficient.
b. A justification of "documentation only" is not acceptable.
5) Risk metrics are not properly calculated.
a. [[ICCDP|ICCDP]] and [[ICLERP|ICLERP]] for TS Completion Time extensions are to be based on the proposed full Completion Time , not the extension, and should include nominal maintenance probabilities of other equipment.
b. ΔCDF and ΔLERF are to be based on reasonable or bounding assumptions on frequency of use of proposed change.
c. Consider that the metrics in the RG assume full scope assessments.
6) Failure to address external events (especially fires) adequately.
a. Commission policy since 2003 has been that significant sources of risk (which could substantially affect the regulatory decision) should be quantified (COMNJD-03-0002).
b. The disposition of out-of-scope initiators as "insignificant" must be convincing, and may require additional [[PRA|PRA]] analyses ([[Regulatory Guide::Regulatory Guide 1.174|RG 1.174]]).
c. If the [[LAR|LAR]] deals with a plant [[SSC|SSC]] which mitigates fires, a quantitative treatment of fire risk will be required.
* This may be a bounding evaluation (using [[IPEEE|IPEEE]] methods).
* The lack of cable route information or other deficiencies is not a justification to ignore the risk source.
* The sole use of fire watches and other compensatory measures without quantitative characterization of risk is not acceptable.
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=APPENDIX F PRE-SUBMITTAL MEETING GUIDANCE=
A pre-submittal meeting is recommended for those licensing actions in which the scope of the action is significant or where substantive changes or new analytical approaches are applied which deviate from existing licensing basis, precedent or current review standards. The timing and extent of the pre-submittal meeting should vary dependent upon the complexity, risk, scope, uniqueness, and cost associated with the proposed licensing action. In some cases, it could be held a year or two prior to the actual submittal.
The following guidance was developed by the NEI Licensing Action Task Force working with the NRC. The full guidance considers a very complex submittal. Licensees should choose the portions of the guidance that are appropriate for the proposed [[LAR|LAR]].
If a pre-submittal meeting is determined to be needed, initial discussions with the NRC Project Manager at the [[LAR|LAR]] conceptual stage are recommended. The discussions should primarily serve to establish periodic communication channels and maintain awareness. These preliminary discussions should also include the [[LAR|LAR]] scope and timing, as well as the identification of NRC branch involvement and key technical or analytical methods. At this early conceptual phase, many of the questions regarding proposed analytical or design approaches cannot be answered, yet a preliminary review will highlight the risk or uncertainty that some of those decisions may incur. Communications at this early stage can also identify opportunities for lead plant or pilot plant applications.
Licensees are cautioned, however, that there are no provisions in the NRC’s regulatory framework that allow the discussions and apparent determinations from a pre-submittal meeting to prejudice or otherwise influence the review outcome of a future submittal. The discussions described above, therefore, are purely guidelines to licensees on how to improve communications with the NRC staff regarding a future licensing action. Anything that transpires before and during a pre-submittal meeting can only be taken as good-faith effort by the NRC technical staff in the absence of a formal licensee submittal.
Internal Licensee Pre-Submittal Preparations
With the [[LAR|LAR]] scope defined and the preliminary regulatory interface/pre-submittal plan established, the next phase is the detailed development of the licensing basis of the submittal, which can be accomplished through the application of the pre-submittal guidance checklist. In support of the checklist, licensees should:
1) Determine the preliminary bases for the proposed change. Within the framework of a licensee's project management and change control processes, licensees should determine:
* The need and the technical basis for the change to the licensing basis or other document or program
- Compliance
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- Cost benefit
* The preliminary development of safety basis for change
- Regulatory requirements
- Operating experience
- Identification, and applicability determination, of precedent
- Consolidated Line-Item Improvement Process and Technical Specification Task Force travelers
* The preliminary determination of safety/security interface
* The preliminary determination of significant hazards determination
* The preliminary determination of environmental consequences
2) Identify precedent and justify applicability to the proposed licensing action. Identify and justify deviations between the approved precedent and the proposed licensing action. For the use of older precedent, provide justification that the precedent meets current NRC regulatory standards.
3) Review the [[SRP|SRP]] and applicable licensing precedent to identify potential technical branch and NRC organizational interfaces for the proposed licensing action. Reconcile the evaluation with the NRC project manager. Provide supporting bases for the conclusions reached, if not self-explanatory.
4) Identify the NRC regulatory guidance, generic communications, facility specific licensing basis and industry guidance and standards that establish the current licensing basis as related to the scope of the proposed licensing action. Identify the proposed changes to each of these documents that will establish the new basis for the proposed licensing action. Evaluate the differences. Similarly evaluate deviations, if any, from the most current versions of the NRC guidance documents. Identify specific computer codes and industry standards to be used, including revision level. Include NRC approval status of any documents. Provide sufficient detail to establish the concise licensing basis of the proposed licensing actions that highlights departures from the current plant licensing basis or current NRC review standards.
The development of the pre-submittal guidance checklist will be iterative. Interface with the NRC project manager after the initial completion to identify the key technical branches and NRC organizations that will be involved in the proposed [[LAR|LAR]] review. This will help to identify required participants in the pre-submittal meeting. If possible, get feedback from the technical branches through the project manager on the key licensing basis assumptions to focus the agenda of the pre-submittal meeting.
Once general alignment is achieved, establish a target date for the pre-submittal meeting ensuring all key technical branches can be represented. The timing of the actual pre-submittal
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meeting will vary, depending on the scope and complexity of the licensing action, as well as the regulatory certainty.
Docketed Pre-Submittal Meeting Letter
The licensee should work with the NRC project manager to determine the timing for submission of pre-submittal meeting materials, including the pre-submittal guidance checklist, to allow the information to be placed in the NRC's [[ADAMS|ADAMS]]. This allows the NRC staff and members of the public participating by teleconference to view the materials prior to the pre-submittal meeting. Ideally, handout materials should be provided prior to the preparation of the meeting announcement, so that the NRC project manager and technical staff have adequate time to review and prepare for the meeting and align on the meeting agenda. Any handout materials developed for the meeting should reflect the information contained in the checklist.
To ensure the meeting is effective, the NRC staff must be provided sufficient time to understand the scope of the request and the complexity of any differences being sought from past reviews prior to the public meeting. To best facilitate this, licensees should consider providing Section I of the pre-licensing checklist sufficiently prior to the meeting to allow NRC review. The meeting slides, which should reflect the information in Section I and Section II of the checklist should be provided at a timeframe agreed upon with the NRC project manager.
The licensee should be clear in the goals and expectations from the meeting, such as alignment on applicable review criteria, guidance on level of detail, justification required for use of a new analytical method, applicability of a precedent, feasibility of a desired schedule, and so forth. Key participants from the licensee and the NRC technical branches should be identified. This assures that the NRC staff has time to review the documents and ensures that all public access goals can be met for noticing meeting materials. Late presentation materials can result in administrative hardships to NRC staff and could jeopardize the effectiveness of the meeting due to its time sensitive nature.
Pre-submittal Meeting
The licensee should work closely with the NRC project manager prior to the pre-submittal meeting to ensure that all necessary staff attends. The licensee should also request the attendance of the branch chief(s) of key NRC technical organization(s) for the submittal. This is particularly important to add continuity if there is a subsequent change in the NRC staff’s reviewer, branch chief or project manager.
The pre-submittal meeting should follow the checklist information to ensure that all key aspects of the licensing action are addressed. The NRC has indicated that it is helpful to identify the plant-specific licensing basis requirements related to the submittal and the specific findings that must be made. The pre-submittal meeting must be thorough and specific in defining the plant specific licensing basis. Use the pre-submittal guidance checklist to present the complete picture of the licensing basis and the proposed bases that support the change.
It is important to engage in dialogue with the NRC staff regarding the regulatory criteria and standards to be applied during the NRC staff’s review of the future licensee submittal. However, during the pre-submittal meeting licensees should take care to not ask questions that seek a
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determination from the NRC on an appropriate course of action. Licensees are reminded that the information provided by the NRC staff during the pre-submittal meeting is non-binding and is not prejudicial to the outcome of a future NRC review.
At the end of the meeting, have a thorough closing in which all parties concur with the results of the meeting. Clearly document the results of the meeting, including any outstanding issues by updating the pre-submittal guidance checklist. Work with the NRC project manager to issue a meeting summary, by resubmitting the revised pre-submittal guidance checklist. Include meeting minutes as applicable. In the subsequent submittal, refer to the pre-submittal meeting and any documented meeting summary.
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ATTACHMENT – PRE-SUBMITTAL GUIDANCE CHECKLIST
Instruction Sheet
SECTION I
Licensee/Plant: Self-explanatory.
Proposed Licensing Action: Provide a short, descriptive title of the proposed licensing action.
Submittal Type Check Boxes: Self-explanatory.
Detailed Description of Proposed Licensing Action: Provide a short paragraph or bulletized list of proposed change(s). Attach additional information as needed. This section should specifically identify the governing regulation, operating license or plant technical specification section affected. The purpose is to provide sufficient detail prior to the meeting for the NRC to review the proposed licensing action prior to the pre-submittal meeting to identify relevant points of agreement/disagreement or areas for discussion. The detail should be sufficient for the NRC to identify the technical branches or organizations that will be involved in the licensing action and the potential pre-submittal meeting.
Planned Submittal Date: Self-explanatory.
Requested Approval Date & Basis: Identify linkage to specific plant conditions or outage and the associated basis.
Topics Addressed in Checklist: Checklist is designed to allow focus on specific aspects of a proposed licensing action. For example, an extended power uprate may need a specific focus on annulus pressurization loads, emergency core cooling operation, and so forth. The contents of this box should be a bulletized list of each aspect that will be contained within this checklist. If the checklist governs the entire proposed licensing action, then this box can be marked not applicable.
Other: Provide any additional information that is necessary for understanding of the proposed licensing action. Summarize any specific considerations as they relate to other regulations such as [[CFR::10 CFR 50.54#a|10 CFR 50.54(a)]], (p) or (q), 50.55a or 50.59. Discuss the potential impact of the proposed licensing action on other ongoing licensing actions, including any potential for consideration as linked licensing actions.
Preliminary Bases for the Proposed Licensing Action: Provide summaries of the applicable bases for the overall proposed licensing action.
* Technical: Summarize the purpose of the proposed change and the associated technical basis.
* Safety: Summarize the preliminary safety bases of the proposed change.
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* Security: Summarize the preliminary security bases of the proposed change.
* No Significant Hazards Consideration: Identify the elements of the analysis that will be key to support the final determination. Identify key assumptions.
* Environmental: Summarize the planned approach (i.e., categorical exclusion, environmental assessment, etc.) and bases for the environmental review. Identify key assumptions.
Regulatory Organization Interface: Review the Standard Review Plan ([[SRP|SRP]]) and applicable licensing precedent to identify potential technical branch and NRC organizational interfaces for the proposed licensing action. Reconcile the evaluation with the NRC project manager. Attach supporting bases for the conclusions reached, if not self-explanatory.
SECTION II
NOTE
A single Section II is developed to address each aspect that is in the checklist. This may lead to several Section IIs included in the checklist.
Proposed Licensing Action: Provide a short, descriptive title of the proposed licensing action.
Topic: The specific topic covered by this section of the checklist as listed in Section I.
Preliminary Bases for the Topic: Provide summaries of the applicable bases for the topic. If any specific area is not needed to support the evaluation of the topic, then it may be deleted.
* Technical: Summarize the purpose of the proposed change and the associated technical basis relevant to the topic.
* Safety: Summarize the preliminary safety bases of the proposed change relevant to the topic.
* Security: Summarize the preliminary security bases of the proposed change relevant to the topic.
* No Significant Hazards Consideration: Identify the elements of the analysis relevant to the topic that will be key to support the final determination. Identify key assumptions.
Precedent: Identify precedent and justify applicability to the proposed licensing action. This list should include any applicable industry requests for additional information that are relevant. Identify and justify deviations between the approved precedent and the proposed licensing action. For the use of older precedent, provide justification that the precedent meets current NRC regulatory standards.
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NOTE
The specific evaluations of the licensing basis of the proposed submittal should be developed early in the conceptual phase of the proposed licensing action. If a pre-submittal meeting is determined to be necessary, the details provided in the following sections will establish the bulk of the agenda for the pre-submittal meeting discussion. Differences between the proposed licensing basis and the current licensing basis and/or the NRC’s current review criteria must be highlighted. Identify specific codes and industry standards to be used, including revision level. Include NRC approval status of any documents. Identify deviations from NRC standards and establish the basis for the proposed deviation. Provide sufficient detail to establish the concise licensing basis of the proposed licensing actions that highlights departures from the current plant licensing basis or current NRC review standards. In some cases, decisions will not have been made regarding specific licensing bases, computer codes, etc. In those cases, at a minimum, the current licensing basis or analytical method should be identified and contrasted to current NRC review standards. Proposed options should be identified. The following sections are broken down in general compliance with regulatory classifications defined in NEI 07-06: The Nuclear Regulatory Process.
NRC Regulatory Guidance and Staff Interpretations: The NRC interprets and clarifies NRC regulatory requirements in generic guidance documents (e.g., NUREGs, regulatory guides, and branch technical positions). NRC guidance is used to communicate approaches acceptable to the NRC staff for meeting NRC requirements. However, as opposed to NRC regulations, the NRC’s regulatory guidance and staff interpretations do not (in and of themselves) have the force of legally binding requirements. Because NRC guidance documents are not the equivalent of NRC rules, the staff interpretations in these documents may be subject to challenge. Methods and solutions different from those set out in the NRC guidance documents may be acceptable to the NRC based upon plant-specific review.
In this section, identify the NRC regulatory guidance that establishes the current licensing basis as related to the scope of the proposed licensing action. Identify the regulatory guidance documents that will be used for the proposed licensing action. Evaluate the differences, if any, between the proposed guidance documents and the guidance documents used in the current licensing basis. Similarly evaluate deviations, if any, from the most current versions of the NRC guidance documents.
NRC Generic Communications: Generic communications address generic concerns that evolve from nuclear reactor operating experience and regulatory initiatives that have broad applicability. These generic communications do not, in and of themselves impose new requirements. However, through the regulatory process may become part of the facility licensing basis.
In this section, identify the NRC generic communications that establish the current licensing basis as related to the scope of the proposed licensing action. Identify the NRC generic communications that will be applied to the proposed licensing action. Evaluate the differences, if any, between the proposed application of the generic communications and the generic communications established in the current licensing basis. Similarly evaluate deviations, if any, from the most current versions of the NRC generic communications.
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Facility Specific Licensing Basis: The facility specific licensing basis is comprised of information exchanged between the licensee and the NRC relating to design features, equipment descriptions, operating practices, site characteristics, programs and procedures, and other factors that describe a plant’s design, construction, maintenance, and operation. Facility specific licensing basis information is contained in a variety of document types (e.g., final safety analysis report, license amendments, regulatory commitments, NRC [[SE|SE]]s and [[SER|SER]]s, etc.).
In this section, identify the facility specific licensing basis documents that establish the current licensing basis as related to the scope of the proposed licensing action. Identify and evaluate any changes to those documents that will be required for the proposed licensing action.
Industry Guidance: Industry guidance in the form of codes and standards, topical reports and other guidance (NEI, [[EPRI|EPRI]], etc.), has historically been developed to address operational, technical and regulatory issues. Application of such industry guidance is typically voluntary. However, specific facility endorsement of these documents can establish part of the licensing basis. Note that these documents will have different levels of industry and NRC, endorsement, review and approval. Some may have specifically been assessed and documented by the NRC in the form of a Safety Evaluation ([[SE|SE]]) report or in a regulatory guide, etc.
Identify the specific industry guidance, codes and standards, etc. that establish the current licensing basis as related to the scope of the proposed licensing action. Identify and evaluate proposed changes. Evaluate the proposed application of industry guidance as it relates to the most current industry guidance. Identify the NRC review and approval level, if any, or past precedent of application.
SECTION III
NOTE
The above pre-submittal guidance checklist information should be developed early in the conceptual phase of a proposed licensing action and can be completed for any proposed licensing action regardless of scope. The completion of the checklist, the associated evaluation and associated dialogue with the NRC project manager will help determine the potential need for, and the timing of, a pre-submittal meeting.
For those licensing actions in which the scope of the action is significant or where substantive changes from existing licensing basis or current review standards are identified, then a pre-submittal meeting is encouraged. If a pre-submittal meeting is determined to be required, then the following sections should be completed.
Meeting Actions/Agreements: Document any additional actions or agreements made in the meeting including any schedule to resolve identified issues.
Pre-submittal Meeting Attendees: All meeting attendees should be identified, specifically identifying the branch or technical organization represented.
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The NRC will issue meeting minutes for public pre-submittal meetings. The NRC should attach this checklist to the meeting minutes or at a minimum ensure that all key points of agreement as noted in this checklist are identified. Licensees should ensure that this pre-submittal guidance checklist is revised to reflect the outcome of the meeting and resubmitted to the NRC for inclusion in the minutes.
Pre-submittal Guidance Checklist – Section I (Provide Section I to NRC [[PM|PM]] approximately 30 days prior to meeting)
Licensee/Plant:
Proposed Licensing Action:
F-10
License Amendment Application
[[CFR::10 CFR 50.55a#|10 CFR 50.55a]] Request
Exemption Request
Other ______________________
Detailed Description of the Proposed Licensing Action:
Planned Submittal Date:
Requested Approval Date & Basis:
Topics Addressed in Checklist:
Other:
Preliminary Bases for the Proposed Licensing Action:
Technical:
Safety:
Security:
No Significant Hazards Consideration:
Environmental:
Pre-submittal Guidance Checklist – Section I (Provide Section I to NRC [[PM|PM]] approximately 30 days prior to meeting)
Licensee/Plant:
Proposed Licensing Action:
F-11
Regulatory Organization Interface:
Reactor Systems
Accident Dose
Security
Electrical
Probabilistic Risk Assessment
Emergency Planning
Instrument and Controls
Fire Protection
Technical Specifications
Mechanical and Civil
Piping and [[NDE|NDE]]
Special Projects
Balance of Plant
Steam Generator Tube Integrity
Health Physics
Containment and Ventilation
Component and Performance Testing
[[DORL|DORL]] - Projects
[[NMSS|NMSS]]
Environmental
[[OE|OE]]
[[NSIR|NSIR]]
[[OGC|OGC]]
Pre-submittal Guidance Checklist – Section II (Provide Section II to NRC [[PM|PM]] for staff review at least 15 days prior to meeting)
Proposed Licensing Action:
Topic:
F-12
Preliminary Bases for the Topic:
Technical:
Safety:
Security:
No Significant Hazards Consideration:
Precedent
Relevance and Deltas to Proposed Licensing Action:
Pre-submittal Guidance Checklist – Section II (Provide Section II to NRC [[PM|PM]] for staff review at least 15 days prior to meeting)
Proposed Licensing Action:
Topic:
F-13
NRC Regulatory Guidance and Staff Interpretations:
Current
Proposed
Comments
Standard Review Plan
Branch Technical Positions
Review Standards
Regulatory Guides
NUREGs
Interim Staff Guidance
Other
Pre-submittal Guidance Checklist – Section II (Provide Section II to NRC [[PM|PM]] for staff review at least 15 days prior to meeting)
Proposed Licensing Action:
Topic:
F-14
NRC Generic Communications:
Current
Proposed
Comments
NRC Bulletins
NRC Generic Letters
Regulatory Information Summaries
NRC Information Notices
Other
Facility Specific Licensing Basis:
Current
Proposed
Comments
[[Licensing Basis Document::Updated Final Safety Analysis Report|Updated Final Safety Analysis Report]]
Technical Specifications Bases
Technical Requirements Manual
Fire Hazards Analysis Report
Pre-submittal Guidance Checklist – Section II (Provide Section II to NRC [[PM|PM]] for staff review at least 15 days prior to meeting)
Proposed Licensing Action:
Topic:
F-15
Facility Specific Licensing Basis:
Current
Proposed
Comments
NRC Safety Evaluation Reports
NRC Correspondence
Regulatory Commitments
Computer Codes
Other
Industry Guidance:
Current
Proposed
Comments
Codes and Standards
Topical Reports
Industry Initiatives and Guidelines
Other
Pre-submittal Guidance Checklist – Section III
Proposed Licensing Action:
Topic:
F-16
Meeting Actions/Agreements:
1.
2.
3.
4.
5.
6.
7.
Pre-submittal Meeting Attendees:
Affiliation/Title:
Contact Information:
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=APPENDIX G VOLUNTARY VS. NON-VOLUNTARY LICENSE AMENDMENT REQUESTS=
In a July 4, 2010 letter from the NEI General Counsel (Ginsberg) to the NRC General Counsel (Burns), (Reference 30), NEI noted that recently issued Regulatory Guides stated, "the methods described in this guide will be used in evaluating compliance with the applicable regulations for license applications, license amendment applications, and amendment requests."
In the NRC's July 14, 2010 response (Reference 31), the NRC stated:
"There are guidance documents which the NRC staff intends only to be 'forward fit,' that is, the guidance will be applied only to: (i) future applicants; and (ii) applications from existing licensees for license amendments, requests for exemptions, and other requests for dispensation from compliance with otherwise-applicable legally binding requirements (an example of such a request would be an application to use an alternative under [[CFR::10 CFR 50.55a#|10 CFR 50.55a)]]. In these circumstances, the NRC does not consider the issuance of 'forward fit' interpretive guidance to constitute 'backfitting.' As the NRC has stated in several different contexts, the Backfit Rule does not protect the expectations of future applicants (including licensees seeking NRC permission to conduct licensed activities in a manner different than what the NRC previously approved) regarding the regulatory requirements that they must meet to obtain NRC approval.
The staff has represented to us that they do not intend to impose on any current nuclear power plant licensee the positions in the three regulatory guides you identified, absent a voluntary request from a licensee to change its licensing basis in a manner which directly implicates the safety issues addressed in those regulatory guides.2 Inasmuch as these regulatory guides are only to be 'forward fit,' the backfitting discussions for those regulatory guides are consistent with the Backfit Rule, and issuance of these regulatory guides do not fall within the purview of 'agency policy' concerning the application of the Backfit Rule to the issuance of interpretive guidance." (Emphasis added.)
Footnote 2 states:
"If a licensee voluntarily seeks to change its licensing basis (i.e., the change is initiated by the licensee to take advantage of a voluntary alternative afforded in the NRC’s regulations, such as the adoption of [[NFPA|NFPA]] 805 under [[CFR::10 CFR 50.48#c|10 CFR 50.48(c)]], and is not compelled by a new or amended regulation), then the NRC may condition its approval of the proposed change upon a licensee agreement to adopt new or revised guidance. Such action will not be deemed to be backfitting if: (i) the new or revised guidance relates directly to the licensee’s voluntary request; and (ii) the specific subject matter of the new or revised guidance is an essential consideration in the NRC staff’s determination of the acceptability of the licensee’s voluntary request."
Most [[LAR|LAR]]s are voluntary and are not required to be submitted. As discussed above, when reviewing a voluntary [[LAR|LAR]], the NRC may condition its approval of the [[LAR|LAR]] on a licensee agreement to adopt new guidance.
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Some [[LAR|LAR]]s are not voluntary and are not subject to the NRC's "forward fit" policy. Examples of such non-voluntary [[LAR|LAR]]s are:
* [[LAR|LAR]]s required for compliance with a new or amended regulation.
* [[LAR|LAR]]s required to respond to a [[CFR::10 CFR 50.54#f|10 CFR 50.54(f)]] letter or Generic Letter.
* [[LAR|LAR]]s to correct non-conservative TS. Administrative Letter 98-10, "Dispositioning of Technical Specifications that are Insufficient to Assure Plant Safety," states that a [[LAR|LAR]] to correct a non-conservative TS must be submitted in a timely fashion.
When submitting a non-voluntary [[LAR|LAR]], it is recommended to state in the cover letter and in the [[LAR|LAR]] justification that the submittal is not voluntary and that the NRC "forward fit" policy does not apply. For example:
"This license amendment request ([[LAR|LAR]]) is required to correct a non-conservative Technical Specification. Currently plant operations are administratively controlled as described in NRC Administrative Letter ([[AL|AL]]) 98-10, 'Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety.' In accordance with the guidance in [[AL|AL]] 98-10, this [[LAR|LAR]] is required to resolve non-conservative TS and is not a voluntary request from a licensee to change its licensing basis. Therefore, this request is not subject to 'forward fit' considerations as described in the letter from S. Burns (NRC) to E. Ginsberg (NEI), dated July 14, 2010 ([[ADAMS|ADAMS]] Accession Number [[document::ML01960180|ML01960180]])."
or
"This license amendment request ([[LAR|LAR]]) is required to comply with [new or amended regulation.] This [[LAR|LAR]] is not a voluntary request from a licensee to change its licensing basis and it is not subject to 'forward fit' considerations as described in the letter from S. Burns (NRC) to E. Ginsberg (NEI), dated July 14, 2010 ([[ADAMS|ADAMS]] Accession Number [[document::ML01960180|ML01960180]])."
Licensees should consider noting that a [[LAR|LAR]] is not voluntary in the letter subject, such as "Non-Voluntary License Amendment Request to Correct a Non-Conservative Technical Specification."
The scope of non-voluntary [[LAR|LAR]]s should be limited to the required issue and not include other voluntary changes.
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=APPENDIX H INDUSTRY CONSOLIDATED AND MULTI-LICENSEE COORDINATED LICENSE AMENDMENT REQUESTS=
NOTE: The following discussion is a placeholder developed by the industry. It will be replaced with a full process description following agreement with the NRC.
There are three [[LAR|LAR]] types:
– Type 1: A plant-specific change for which there is no NRC-accepted model application;
– Type 2: A change based on an NRC-approved generic justification, such as a Technical Specifications Task Force ([[TSTF|TSTF]]) traveler, for which there is an NRC-accepted model application that requires submittal of technical, plant-specific information; and
– Type 3: A change based on an NRC-approved generic justification, such as a [[TSTF|TSTF]] traveler, for which there is an NRC-accepted model application that does not require submittal of technical, plant-specific information other than choosing the appropriate plant specific options provided in the model application.
The industry has proposed three [[LAR|LAR]] submittal and review processes to the NRC. The Industry Coordinated and Multi-Licensee Consolidated [[LAR|LAR]] processes are new concepts and will require further discussion and agreement with the NRC. The concepts will be expanded and [[LAR|LAR]] directions provided in a future revision to [[NEI::NEI 06-02|NEI 06-02]].
– Plant Specific: The current plant-specific [[LAR|LAR]] process, in which an individual licensee researches, drafts, and submits a [[LAR|LAR]] to the NRC for review. This process is the most encompassing, plant-specific, and resource intensive review process. This process may be used for Type 1, 2, and 3 [[LAR|LAR]] types.
– Industry Coordinated: A [[LAR|LAR]] process in which a group of licensees coordinate the research, drafting, submittal, and NRC reviews. This ensures the [[LAR|LAR]]s have similar content and format and, preferably, the same NRC reviewers. This process is intended to reduce the required industry and NRC resources needed to submit and approve a [[LAR|LAR]]. The Industry Coordinated [[LAR|LAR]] process is recommended for Type 2 or 3 [[LAR|LAR]]s.
– Multi-Licensee Consolidated: A [[LAR|LAR]] process in which a single [[LAR|LAR]] is submitted by a group of licensees (similar to a fleet submittal except the submittal would apply to multiple licensees). This process is intended to be used for [[LAR|LAR]]s that require minimal plant specific technical information and have an NRC accepted model application. The submittal will contain an attachment for each plant containing the authorized signature, a description of any differences, and the TS markups. This process would allow one NRC review for multiple licensees and plants, significantly reducing industry submittal and NRC review resources. The Multi-Licensee Consolidated [[LAR|LAR]] process is recommended for Type 3 [[LAR|LAR]]s.
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Industry Coordinated – The Industry Coordinated [[LAR|LAR]] Process
The Industry Coordinated [[LAR|LAR]] Process will be used when a group of licensees coordinate research, drafting, submittal, and NRC review of their submittals. The coordination would typically include providing an industry submittal schedule to the NRC and obtaining feedback on timing and sequencing, which, coupled with consistent content and format, will increase the efficiency of the NRC review. The Industry Coordinated [[LAR|LAR]] process is recommended for Type 2 or 3 [[LAR|LAR]]s.
When using the Industry Coordinated [[LAR|LAR]] Process, each [[LAR|LAR]] will be prepared consistent with the NRC-accepted model application in accordance with [[NEI::NEI 06-02|NEI 06-02]]. The difference between the Industry Coordinated [[LAR|LAR]] Process and the Plant-Specific [[LAR|LAR]] Process is the group scheduling of [[LAR|LAR]] submittals and NRC reviews. The Industry Coordinated [[LAR|LAR]] Process would seek to use the same NRC reviewer (or group of reviewers) to review [[LAR|LAR]]s submitted to the NRC on a scheduled, coordinated fashion.
Multi-Licensee Consolidated – The Multi-Licensee Consolidated [[LAR|LAR]] Process
The Multi-Licensee Consolidated [[LAR|LAR]] Process will be used to submit a single [[LAR|LAR]] for multiple licensees. This process would be applied to simple, generic changes with little plant-specific information required. The Multi-Licensee Consolidated Submittal Process is recommended for Type 3 [[LAR|LAR]]s.
The Multi-Licensee Consolidated [[LAR|LAR]] would include:
* A common cover letter from the sponsoring organization ([[TSTF|TSTF]], NEI, Owners Group, etc.)
* An enclosure containing the NRC-accepted model application
* Separate attachments for each participating plant. Each attachment would include:
o A statement that the generic justification has been reviewed and is applicable taken from the model application
o A discussion of any variations from the approved justification
o The licensee-authorized signature under oath and affirmation or unsworn declaration
o TS markups
o TS retyped pages (optional)
o TS Bases markups (typically for information only)
The sponsoring organization would prepare and/or assemble the [[LAR|LAR]] with plant-specific data provided and/or verified by each participating licensee. Some steps in the general [[LAR|LAR]] process, such as identification of implementation actions, may be skipped or performed after [[LAR|LAR]] submittal.
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Even though the [[LAR|LAR]] would be assembled by an outside party, the licensee must still ensure the submittal meets all regulatory and procedural requirements. For example:
* All plants will need to ensure that the preparation, peer review, verification, technical review, management review, and safety committee review of the [[LAR|LAR]] is performed in accordance with their procedures.
* Change impacts must be identified, tracked, and implemented.
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=APPENDIX I GLOSSARY=
ACCEPTANCE REVIEW
When the holder of a Part 50 license to operate a commercial nuclear power plant submits a REQUEST [[FOR|FOR]] LICENSING ACTION ([[RLA|RLA]]), the NRC staff conducts an ACCEPTANCE REVIEW in accordance with [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-109|LIC-109]]. The ACCEPTANCE REVIEW is the staff’s initial determination of whether the [[RLA|RLA]] reasonably appears to contain sufficient technical information, both in scope and depth, for the staff to complete a detailed technical review and render, in an appropriate time frame for the associated action, an independent assessment of the proposed action with regard to applicable regulatory requirements and the protection of public health, safety, and security. An [[RLA|RLA]] that was rejected during the acceptance review may be resubmitted if is supplemented to address the reasons for the NRC staff’s rejection. If an [[RLA|RLA]] is rejected, the NRC does not issue a No Significant Hazards Consideration Determination (NSHCD) and there are no subsequent hearing rights.
ADEQUATE PROTECTION
The Atomic Energy Act (Ref. 15) delegates to NRC the responsibility to interpret what is necessary to meet the ADEQUATE PROTECTION standard. NRC establishes what is meant by ADEQUATE PROTECTION through rulemaking and the adjudicatory process. In general, ADEQUATE PROTECTION is presumptively assured by compliance with NRC requirements. The NRC staff evaluates situations of noncompliance to determine the degree of risk and whether immediate action is necessary. If the NRC determines that non-compliance itself is of such safety significance that ADEQUATE PROTECTION is no longer provided, or that it was caused by a deficiency so significant it questions a licensee’s ability to ensure ADEQUATE PROTECTION, the NRC may demand immediate action, up to and including shutdown or cessation of licensed activities.
APPLICABLE STAFF POSITION
An APPLICABLE STAFF POSITION ([[ASP|ASP]]) (Ref. 16) is a formal NRC staff position documented in writing in the [[SRP|SRP]], a Branch Technical Position ([[BTP|BTP]]), a RG, or an NRC staff [[SE|SE]] to which the staff or a licensee/applicant has previously committed or relied upon as documented in the CURRENT LICENSING BASIS. This includes staff positions taken in response to a TOPICAL REPORT from an Owners Group, NEI, [[EPRI|EPRI]], or [[NSSS|NSSS]] vendor. For a given issue, the most recent [[ASP|ASP]] is the appropriate baseline for use in subsequent plant-specific licensing actions pertaining to the regulatory issue or requirement in question.
BACKFIT
The Commission recognized the importance of BACKFIT controls in 1988 when it approved a change to [[CFR::10 CFR 50.109#|10 CFR 50.109]] to establish administrative standards for NRC imposition of new regulations or new interpretations of existing regulations. The rule defines a BACKFIT as "the modification of or addition to systems, structures, components, or design of a facility; or the
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design approval or manufacturing license for a facility; or the procedures or organization required to design, construct or operate a facility; any of which may result from a new or amended provision in the Commission's regulations or the imposition of a regulatory staff position interpreting the Commission's regulations that is either new or different from a previously applicable staff position …" (Ref. 17).
COMMITMENT
See Regulatory Commitment.
CURRENT LICENSING BASIS
The term CURRENT LICENSING BASIS ([[CLB|CLB]]) is not defined in [[CFR::10 CFR 50#|10 CFR 50]]. However, the following is a practical definition of the [[CLB|CLB]] that is derived from [[CFR::10 CFR 54#|10 CFR 54]]. The [[CLB|CLB]] for an operating commercial reactor is comprised of:
The set of NRC regulations applicable to a specific plant that are docketed and in effect (e.g., [[CFR::10 CFR 2#|10 CFR Parts 2,]] 19, 20, 21, 26, 30, 40, 50, 51, 52, 54, 55, 70, 72, 73, 100 and appendices thereto).
The set of written licensee commitments for ensuring compliance with and operation within applicable NRC requirements and the plant-specific design basis (including all modifications and additions to such commitments over the life of the license) that are docketed and in effect. This includes formal commitments contained in docketed licensing correspondence (e.g., licensee event reports and responses to NRC bulletins, generic letters, and enforcement actions) and NRC [[SE|SE]]s. See OBLIGATION (Ref. 3).
The plant-specific design bases information defined in [[CFR::10 CFR 50.2#|10 CFR 50.2]] and documented in the plant-specific Final Safety Analysis Report ([[FSAR|FSAR]]) as updated in accordance with [[CFR::10 CFR 50.71#e|10 CFR 50.71(e)]].
COMPLIANCE
The term COMPLIANCE means that a [[SSC|SSC]] satisfies all requirements of applicable rules, regulations, orders, and licenses (including TS). COMPLIANCE is based on the intent of the requirement at the time of its promulgation. The NRC typically documents the intent of a requirement in a Federal Register Notice, and licensees typically incorporate implementing language into the [[CLB|CLB]] by updating the [[FSAR|FSAR]] or other licensee-controlled document. NRC regulations ([[CFR::10 CFR 50.59#|10 CFR 50.59]] and [[CFR::10 CFR 50.109#|10 CFR 50.109)]], supplemented by NRC and licensee procedures, control the imposition of new or different interpretations.
DENIAL
DENIAL of a [[LAR|LAR]] can occur after it has been accepted by NRC for review. In accordance with [[CFR::10 CFR 2.108#|10 CFR 2.108,]] NRC may deny a [[LAR|LAR]] if the applicant fails to respond to a [[RAI|RAI]] within thirty (30) days from the date of the request, or within such other time as may be specified. Notices of [[LAR|LAR]] DENIAL are posted in the Federal Register. [[LAR|LAR]]s that have been denied cannot be resubmitted without substantial revision to address the reason(s) for DENIAL. Also, a hearing may be
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requested because a No Significant Hazards Consideration Determination will have been published by the NRC following initial acceptance of the [[LAR|LAR]] for staff review.
DESIGN BASIS
The plant-specific DESIGN BASIS, per [[CFR::10 CFR 50.2#|10 CFR 50.2,]] is the set of information that identifies the specific functions to be performed by a facility’s [[SSC|SSC]]s, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values may be (1) restraints derived from generally accepted state-of-the-art practices for achieving functional goals, or (2) requirements derived from analysis (based on calculation and/or experiments) of the effects of a postulated accident for which a [[SSC|SSC]] must meet its functional goals. DESIGN BASIS program guidelines are discussed in [[NEI::NEI 97-04|NEI 97-04]] (Ref. 1) which is endorsed by [[Regulatory Guide::Regulatory Guide 1.186|Regulatory Guide 1.186]].
DETERMINISTIC
The term DETERMINISTIC is used to differentiate prescriptive requirements from those that are "risk informed." The NRC initially developed many of its regulations without considering numerical estimates of risk. They are "deterministic" in the sense that they set strict limits or prescribe discrete outcomes. DETERMINISTIC requirements are established based on experience, test results, and expert judgment considering factors such as design margin, defense-in-depth, and accident prevention or mitigation. Compare with PROBABILISTIC.
DURABLE GUIDANCE
The term DURABLE GUIDANCE refers to the type of regulatory document that is used to disseminate an APPLICABLE STAFF POSITION. A regulatory guidance document is "durable" if changes to the document are controlled by an administrative approval process that includes an opportunity for the public to comment on proposed changes. For example, RGs, the [[SRP|SRP]], NRC [[SE|SE]]s, and the Statements of Consideration that accompany Final Rules are DURABLE GUIDANCE documents.
EMERGENCY LICENSE AMENDMENT
The term EMERGENCY LICENSE AMENDMENT (Ref. 18) applies to situations where the Commission finds that failure to act in a timely way would result in de-rating or shutdown of a nuclear power plant, or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level. In such cases the NRC may issue a license amendment involving a NSHCD without prior notice and opportunity for hearing or for public comment.
EXIGENT LICENSE AMENDMENT
The term EXIGENT LICENSE AMENDMENT (Ref. 18) applies to situations where a [[LAR|LAR]] is submitted with a need date of more than seven days but less than four or five weeks from submittal. The preferred exigent process is to use a shortened public notice period in the Federal Register. Local media may be used to notice [[LAR|LAR]]s that require disposition in less time than needed for a 2-week comment period in the Federal Register.
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FINDING
See REGULATORY FINDING.
FIRST OF A KIND ([[FOAK|FOAK]]) LICENSE AMENDMENT
The term FIRST OF A KIND ([[FOAK|FOAK]]) applies to [[LAR|LAR]]s that involve new or more complex technology, a greater scope of applicability, or a greater organizational complexity than previously reviewed by the NRC staff. There is no close existing analogue or body of precedent that staff can use to aid the regulatory review of an [[FOAK|FOAK]] [[LAR|LAR]]. Consequently, the NRC review fees for processing such an application can be quite high and pre-submittal meetings are important for the ACCEPTANCE REVIEW.
FLEET [[LAR|LAR]]
The term FLEET [[LAR|LAR]] refers to a [[LAR|LAR]] that is submitted by an operating company on behalf of several nuclear units (i.e., more than one Docket). The NRC recommends that separate FLEET [[LAR|LAR]] submittals should be provided for [[ITS|ITS]] and non-ITS plants.
GENERIC
The term GENERIC refers to all members of a genus, class, group, or kind (Ref. 19). In the context of nuclear power, it pertains to all of power plants or organizations that have common characteristics, such as the set of [[BWR|BWR]]s, the set of [[PWR|PWR]]s, the set of reactors designed by a particular [[NSSS|NSSS]] vendor, etc.
GENERIC ISSUE
In its broadest sense, a GENERIC ISSUE is a well-defined, discrete issue with the following attributes:
* The issue could affect public health and safety, the common defense and security, or the environment
* The issue applies to two or more facilities, licensees, certificate holders, or holders of other regulatory approvals
* The issue is not being addressed by an existing program or process
* The issue cannot be readily addressed by existing regulations, policies, guidance, or voluntary industry initiatives
* The issue cannot be readily addressed by other regulatory programs or processes
* The risk or safety significance of the issue can be adequately determined
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A GENERIC ISSUE may lead to regulatory changes that either enhance safety, or reduce unnecessary regulatory burden.
GENERIC SAFETY ISSUE
A regulatory issue is a GENERIC SAFETY ISSUE if it falls into one of five groups defined in [[NUREG::NUREG-0933|NUREG-0933]], "A Prioritization of Generic Safety Issues:" (Ref. 20)
* [[TMI|TMI]] Action Plan items ([[NUREG::NUREG-0660|NUREG-0660]] (Ref. 21) and [[NUREG::NUREG-0737|NUREG-0737]] (Ref. 22)
* Task Action Plan items ([[NUREG::NUREG-0371|NUREG-0371]], (Ref. 23) [[NUREG::NUREG-0471|NUREG-0471]], (Ref. 24) and subsequent unresolved safety issues)
* New generic issues identified from various sources
* Human factors issues ([[NUREG::NUREG-0985|NUREG-0985]] (Ref. 25)
* Chernobyl issues ([[NUREG::NUREG-1251|NUREG-1251]](Ref. 26)
LICENSE AMENDMENT REQUEST
A LICENSE AMENDMENT REQUEST ([[LAR|LAR]]) is a formal request from a licensee to amend a Part 50 facility operating license pursuant to [[CFR::10 CFR 50.90#|10 CFR 50.90]] (application for amendment of license, construction permit, or early site permit), 50.91 (notice for public comment; state consultation), and 50.92 (issuance of amendment). A licensee submits a [[LAR|LAR]] whenever it determines that a proposed activity (e.g., plant modification, procedure change) requires modification of the plant [[OL|OL]] or TS.
LINKED SUBMITTAL
A [[LAR|LAR]] that relies upon another [[LAR|LAR]] which is under review by NRC and not yet approved.
MODEL SAFETY EVALUATION
A MODEL SAFETY EVALUATION is prepared by the NRC staff for an approved [[TSTF|TSTF]] traveler.
OBLIGATION
An OBLIGATION is any condition or action that is a legally binding requirement imposed on licensees through applicable rules, regulations, orders and licenses (including TS and license conditions). These conditions (also referred to as regulatory requirements) generally require formal NRC approval as part of the change-control process. Also included in the category of OBLIGATIONs are those regulations and license conditions that define change-control processes and reporting requirements for licensing basis documents such as the Updated [[FSAR|FSAR]], Quality Assurance Program, Emergency Plan, Security Plan, and Fire Protection Program.
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PILOT PLANT (for TOPICAL REPORTS or [[FOAK|FOAK]] [[LAR|LAR]]s)
A PILOT PLANT (as distinguished from a Lead Plant) is a licensee that submits a [[LAR|LAR]] for a review by NRC that will result in a plant-specific license amendment for the pilot plant, and also lead the way for additional plants to submit similar [[LAR|LAR]]s. The NRC will consider accepting a PILOT PLANT [[LAR|LAR]] if it will assist the staff in identifying enhancements to the NRC’s generic regulatory program by identifying process improvements and lessons learned for review of a future [[LAR|LAR]].
PRECEDENT
The term PRECEDENT refers to prior NRC licensing actions and decisions which may provide additional clarification for the acceptability of a proposed licensing action. In and of itself, PRECEDENT action does not provide justification for approval of a proposed license amendment; however, effective discussion of PRECEDENT can assist in the regulatory review by presenting technical and regulatory considerations applied to prior similar licensing actions. The effective use of PRECEDENT depends to a large extent on the level of detail provided in the applicant’s [[LAR|LAR]] with respect to similarities and differences between the PRECEDENT and the proposed licensing action.
PROBABILISTIC
The term PROBABILISTIC is associated with a systematic analysis for addressing risk as it relates to the performance of complex systems to understand likely outcomes, sensitivities, areas of importance, system interactions, and areas of uncertainty (Ref. 27). Compare with DETERMINISTIC.
REGULATORY ANALYSIS
The NRC has prepared and published guidance on the performance of a REGULATORY ANALYSIS to ensure sound decisions regarding actions needed to protect the health and safety of the public or the common defense and security. Regulatory analyses are required for all regulatory actions that involve BACKFITTING.
REGULATORY COMMITMENT
A REGULATORY COMMITMENT (Refs. 3 and 4) is an explicit statement to take a specific action agreed to, or volunteered by, a licensee and submitted to NRC in writing on the licensee’s docket.
REGULATORY FINDING
The term REGULATORY FINDING refers to the NRC staff’s explanation and bases, documented in a written SAFETY EVALUATION in response to a REQUEST [[FOR|FOR]] LICENSING ACTION, for concluding 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) such activities will be conducted in compliance with the Commission’s regulations, and (3) the
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issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
REGULATORY MARGIN
REGULATORY MARGIN is a subjective concept analogous to equity, or good will. It represents a licensee’s intangible reputation based on past performance. Some observers perceive that strong licensee performance (e.g., a good record of compliance, a good reputation for quality submittals, fewer operational problems, cordial professional relationships among peers, etc.) leads to greater REGULATORY MARGIN.
REQUEST [[FOR|FOR]] ADDITIONAL INFORMATION ([[RAI|RAI]])
An NRC REQUEST [[FOR|FOR]] INFORMATION ([[RAI|RAI]]) is either a formal or an informal request for information needed by the NRC staff to document the basis for the staff’s conclusions in the [[SE|SE]] of a REQUEST [[FOR|FOR]] LICENSING ACTION.
REQUEST [[FOR|FOR]] LICENSING ACTION
A REQUEST [[FOR|FOR]] LICENSING ACTION ([[RLA|RLA]]) is a formal request from a licensee/applicant that requires prior NRC approval before it can be implemented. A LICENSE AMENDMENT REQUEST ([[LAR|LAR]]) is a special case of an [[RLA|RLA]].
REQUIREMENT
The term REQUIREMENT is a legally binding statute, regulation, license condition, TS, or order. See OBLIGATION.
RISK-INFORMED
A [[LAR|LAR]] is RISK-INFORMED (Ref. 28) if it applies quantitative or qualitative risk insights and techniques in accordance with applicable regulatory guidance. Since 1975, the NRC and its licensees have advanced their knowledge of (and experience with) Probabilistic Risk Assessment ([[PRA|PRA]]). [[PRA|PRA]] considers nuclear safety in a comprehensive way by examining the likelihood of a broad spectrum of initiating events.
SAFETY EVALUATION
The results of the NRC staff’s evaluation of a [[LAR|LAR]] are documented in a SAFETY EVALUATION ([[SE|SE]]). The [[SE|SE]] describes the staff’s technical and regulatory evaluation with respect to the impact on public health and safety of operation in the manner proposed in the [[LAR|LAR]].
SUBMITTAL QUALITY
SUBMITTAL QUALITY depends on the extent to which the submitter understands a set of objective and subjective expectations (e.g., format, content, scope, level of detail, etc.), and the degree to which the submitter can conclude that that a submittal meets these expectations. High-
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quality submittals optimize the presentation of relevant information (e.g., system descriptions, results of calculations, bases that support compliance with applicable requirements, bases that support conformance with applicable NRC and industry guidance, comparisons with precedent, references, definitions, procedures, commitments, implementation plans and schedules, etc.).
TASK INTERFACE AGREEMENT ([[TIA|TIA]])
The term TASK INTERFACE AGREEMENT ([[TIA|TIA]]) refers to an internal process used by [[NRR|NRR]] to respond to requests for technical assistance from an NRC Region or another NRC office. A [[TIA|TIA]] contains questions on subjects involving regulatory or policy interpretations, specific plant events, or inspection findings within the scope of the [[NRR|NRR]] mission and responsibilities. The requesting organization may use a [[TIA|TIA]] to obtain information on specific plant licensing basis, applicable staff positions for an issue, regulatory requirements, [[NRR|NRR]] technical positions, or the safety or risk significance of particular plant configurations or operating practices. The [[TIA|TIA]] process is designed to ensure that concerns from the regions and other NRC offices are resolved in a timely manner and that the [[NRR|NRR]] responses are appropriately communicated. The process includes a provision to interact with licensees to ensure clear and accurate information.
TECHNICAL SPECIFICATION TASK FORCE ([[TSTF|TSTF]]) TRAVELER
The [[TSTF|TSTF]], in support of the [[PWR|PWR]], [[BWR|BWR]], and AP1000® Owners Groups, develops generic changes to the [[STS|STS]]. The documentation in support of the change is called a [[TSTF|TSTF]] traveler. Following approval by the NRC, the traveler number is given an "-A" suffix to indicate that it is approved. Travelers that are approved by the [[TSTF|TSTF]] for industry use as a template and not submitted to the NRC for review are given a "-T" suffix.
TOPICAL REPORT
The term TOPICAL REPORT ([[TR|TR]]) refers to a wide variety of stand-alone reports that are submitted for NRC review. Normally these reports contain technical information about a nuclear power plant safety issue, although they can address policy and process issues also. [[TR|TR]]s improve the efficiency of the licensing process by establishing generic methodologies, designs, procedures, or other guidance that multiple licensees can reference in plant-specific [[RLA|RLA]]s. Internal NRC guidance for reviewing [[TR|TR]]s is contained in [[NRC LIC::LIC-500|LIC-500]] (Ref. 29)
WITHDRAWAL
The term WITHDRAWAL applies at any point after a [[LAR|LAR]] is submitted, but before the NRC staff has completed its regulatory review. A licensee may withdraw a [[LAR|LAR]] without prejudice (i.e., it can be resubmitted at a later date).
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=APPENDIX J [[Acronym|ACRONYM]]S=
[[ADAMS|ADAMS]] Agency Documents Access and Management System
[[APOG|APOG]] AP1000® Owners Group
[[ASME|ASME]] American Society of Mechanical Engineers
[[BWR|BWR]] Boiling Water Reactor
[[BWROG|BWROG]] Boiling Water Reactors Owners’ Group
[[CDF|CDF]] Core Damage Frequency
CFR Code of Federal Regulations
[[CLB|CLB]] Current Licensing Basis
[[CLIIP|CLIIP]] Consolidated Line Item Improvement Process
[[EPRI|EPRI]] Electric Power Research Institute
[[FOAK|FOAK]] First of a Kind
[[FSAR|FSAR]] Final Safety Analysis Report
GDC General Design Criteria
[[INPO|INPO]] Institute of Nuclear Plant Operations
[[ISTS|ISTS]] Improved Standard Technical Specifications
[[LAR|LAR]] License Amendment Request
LCO Limiting Condition for Operation
[[LERF|LERF]] Large Early Release Frequency
NEI Nuclear Energy Institute
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NRC Nuclear Regulatory Commission
[[NRR|NRR]] Office of Nuclear Reactor Regulation
[[NSSS|NSSS]] Nuclear Steam System Supplier
[[OGC|OGC]] Office of the General Counsel
[[OL|OL]] Operating License
[[PM|PM]] (NRC) Project Manager
[[PWR|PWR]] Pressurized Water Reactor
[[PWROG|PWROG]] Pressurized Water Reactor Owners Group
[[RAI|RAI]] Request for Additional Information
[[RLA|RLA]] Request for Licensing Activity (NRC term for a [[LAR|LAR]])
[[RIS|RIS]] Regulatory Issue Summary
[[SE|SE]] Safety Evaluation
[[SOC|SOC]] Statements of Consideration
[[SRP|SRP]] Standard Review Plan
[[SSC|SSC]] Structure, System, or Component
[[TIA|TIA]] Task Interface Agreement
TS Technical Specification
[[TSTF|TSTF]] Technical Specification Task Force
[[UFSAR|UFSAR]] [[Licensing Basis Document::Updated Final Safety Analysis Report|Updated Final Safety Analysis Report]]
USC U.S. Code
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=APPENDIX K REFERENCES=
1. [[NEI::NEI 96-07|NEI 96-07]], Revision 1, "Guidelines for [[CFR::10 CFR 50.59#|10 CFR 50.59]] Implementation," (November 2000).
2. NRC [[Regulatory Guide::Regulatory Guide 1.187|Regulatory Guide 1.187]], "Guidance for Implementation of [[CFR::10 CFR 50.59#|10 CFR 50.59]]; Changes, Tests, and Experiments," (November 2000).
3. [[NEI::NEI 99-04|NEI 99-04]], "Guidelines for Managing NRC Commitment Changes," (July 1999).
4. NRC [[Regulatory Issue Summary::RIS 2000-17|Regulatory Issue Summary 2000-17]], "Managing Regulatory Commitments Made by Power Reactor Licensees to the NRC Staff," (September 21, 2000).
5. [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-100|LIC-100]], "Control of Licensing Bases for Operating Reactors."
6. [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-101|LIC-101]], "License Amendment Review Procedures."
7. [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-109|LIC-109]], "Acceptance Review Procedures."
8. NRC [[Regulatory Issue Summary::RIS 2001-22|Regulatory Issue Summary 2001-22]], "Attributes of a Proposed No Significant Hazards Consideration."
9. [[CFR::10 CFR 51.22#|10 CFR 51.22,]] "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review."
10. [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-105|LIC-105]], "Managing Regulatory Commitments Made By Licensees to the NRC"
11. TSTF-GG-05-01, "Writer's Guide for Plant-Specific Improved Technical Specifications," June 2005.
12. TSTF-GG-05-01, "Writer's Guide for Plant-Specific Improved Technical Specifications," June 2005.
13. NRC [[Regulatory Issue Summary::RIS 2007-06|Regulatory Issue Summary 2007-06]], "[[Regulatory Guide::Regulatory Guide 1.200|Regulatory Guide 1.200]] Implementation," (March 2007).
14. [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-600|LIC-600]], "Changing the Standard Technical Specifications by means of Technical Specifications Task Force ([[TSTF|TSTF]]) Travelers."
15. U.S. Code, [[Atomic Energy Act of 1954|Atomic Energy Act of 1954]] (as amended), Section 182, "License Applications."
16. NRC Management Directive [[MD|MD]] 8.4, "NRC Program for Management of Plant-Specific Backfitting of Nuclear Power Plants."
17. [[CFR::10 CFR 50.109#a1|10 CFR 50.109(a)(1)]], "Backfitting" (definition).
18. [[CFR::10 CFR 50.91#|10 CFR 50.91,]] "Notice for public comment; State consultation."
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19. Dictionary.com (http://dictionary.reference.com/browse/generic).
20. [[NUREG::NUREG-0933|NUREG-0933]], "A Prioritization of Generic Safety Issues."
21. NRC [[NUREG::NUREG-0660|NUREG-0660]], Revision 1, "TMI-2 Action Plan," (August 1980).
22. NRC [[NUREG::NUREG-0737|NUREG-0737]], "Clarification of [[TMI|TMI]] Action Plan Requirements," (November 1980).
23. NRC [[NUREG::NUREG-0371|NUREG-0371]], "Task Action Plans for Generic Activities (Category A)," (November 1978).
24. NRC [[NUREG::NUREG-0471|NUREG-0471]], "Generic Task Problem Descriptions (Categories B, C, and D)," (June 1978).
25. NRC [[NUREG::NUREG-0985|NUREG-0985]], Revision 1, "Human Factors Program Plan," (April 1986).
26. NRC [[NUREG::NUREG-1251|NUREG-1251]], "Implications of the Accident at Chernobyl for Safety Regulation of Commercial Nuclear Power Plants in the United States," (April 1989).
27. NRC Glossary (http://www.nrc.gov/reading-rm/basic-ref/glossary/probabilistic-risk-analysis.html).
28. NRC [[Regulatory Guide::Regulatory Guide 1.174|Regulatory Guide 1.174]], Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," (November 2002).
29. [[NRR|NRR]] Office Instruction [[NRC LIC::LIC-500|LIC-500]], "Topical Report Process."
30. Letter from E. Ginsberg (NEI) to S. Burns (NRC), dated June 4, 2010. [[ADAMS|ADAMS]] Accession Number [[document::ML101970353|ML101970353]].
31. Letter from S. Burns (NRC) to E. Ginsberg (NEI), dated July 14, 2010, [[ADAMS|ADAMS]] Accession Number [[document::ML01960180|ML01960180]].