ML22252A142

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NRR E-mail Capture - NRC Comments on Technology Inclusive Risk Informed Change Evaluation (Tirice) for Non-Light Water Reactors Draft Guidance Document
ML22252A142
Person / Time
Issue date: 09/07/2022
From: Michael Orenak
NRC/NRR/DANU
To: Chisholm B, Tschiltz M
Southern Company Services
References
SOW-16166 Contract No. 221666, SC-16166-108 Rev D
Download: ML22252A142 (72)
v • d • e


Text

From: Orenak, Michael Sent: Wednesday, September 7, 2022 5:38 PM To: michael.tschiltz@gmail.com; Chisholm, Brandon Michael Cc: Sebrosky, Joseph

Subject:

NRC Comments on TIRICE draft guidance document.

Attachments: TIRICE Guidance DRAFT Rev D_final_NRC Comments_CLEAN.docx Brandon and Mike, Attached is the NRCs comments on the TIRICE draft guidance document for discussion in the 9/13 meeting. These will also be available on our public meeting notice site for the 9/13 meeting for public review.

I will be out of the office until 9/13, so please contact Joe Sebrosky with any questions in the meantime.

Mike Michael D. Orenak, Project Manager Advanced Reactor Licensing Branch 1 (UAL1)

NRR - Division of Advanced Reactors and Non-Power Production and Utilization Facilities (DANU) 301-415-3229 Michael.Orenak@nrc.gov

Hearing Identifier: NRR_DRMA Email Number: 1772 Mail Envelope Properties (SA1PR09MB8784F22C900049A97B9D462E80419)

Subject:

NRC Comments on TIRICE draft guidance document.

Sent Date: 9/7/2022 5:37:37 PM Received Date: 9/7/2022 5:37:41 PM From: Orenak, Michael Created By: Michael.Orenak@nrc.gov Recipients:

"Sebrosky, Joseph" <Joseph.Sebrosky@nrc.gov>

Tracking Status: None "michael.tschiltz@gmail.com" <michael.tschiltz@gmail.com>

Tracking Status: None "Chisholm, Brandon Michael" <BMCHISHO@SOUTHERNCO.COM>

Tracking Status: None Post Office: SA1PR09MB8784.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 616 9/7/2022 5:37:41 PM TIRICE Guidance DRAFT Rev D_final_NRC Comments_CLEAN.docx 1008330 Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:

Style Definition: Centered Technology Inclusive Risk Informed Change Evaluation (TIRICE)

For Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 Draft Report Revision D Issued for Collaborative Review Document Number SC-16166-108 Rev D Battelle Energy Alliance, LLC Contract No. 221666 SOW-16166 August 2022 Prepared for:

U.S. Department of Energy (DOE)

Office of Nuclear Energy Under DOE Idaho Operations Office Contract DE-AC07-05ID14517

Technology Inclusive Risk Informed Change Evaluation (TIRICE)

For Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 Draft Report Revision D Document Number SC-16166-108 Rev D Battelle Energy Alliance, LLC Contract No. 221666 SOW-16166 Issued for Collaborative Review by:

Brandon M. Chisholm, Research Engineer Date Southern Company Services

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 Disclaimer This report was prepared as an account of work sponsored by an agency of the United States (U.S.) Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, nor Southern Company, Inc., nor any of its employees, nor any of its subcontractors, nor any of its sponsors or co-funders, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the U.S. Government or any agency thereof. The views and opinions of the authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof.

iii

Foreword Nuclear Regulatory Commission (NRC) regulation 10 CFR 50.59 establishes criteria for determining if prior NRC approval is required before implementing changes to a reactor licensed under 10 CFR Part 50 or 10 CFR Part 52. Nuclear Energy Institute (NEI) document NEI 96-07 Guidelines for 10 CFR 50.59 Implementation provides guidance for applying the 10 CFR 50.59 criteria to currently operating light water reactors (LWRs). This guidance was developed based on the existing change control guidance in NEI 96-07, with appropriate additions and adjustments as provided herein. This guidance provides an alternative change evaluation process for determining if NRC approval is required before implementing certain facility changes to reactors that were licensed using the methodologies in NEI 18-04 Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development and NEI 21-07 Technology Inclusive Guidance for Non-Light Water Reactors -

Safety Analysis Report Content for Applicants Using the NEI 18-04 Methodology. Used in conjunction with an enabling license condition and a full, or partial, exemption to 10 CFR 50.59, this guidance allows reactor licensees that have utilized this approach to implement appropriate change control programs for the operation of their reactors

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 Table of Contents Disclaimer .............................................................................................................................................. iii Foreword ............................................................................................................................................... iv List of Abbreviations ...............................................................................................................................vii

1. Introduction..................................................................................................................................... 1 1.1. Background ............................................................................................................................ 1 1.2. Application of this Guidance................................................................................................... 3 1.3. Purpose and Scope ............................................................................................................... 43 1.4. Relationship of this Guidance to Regulatory Requirements and Controls ................................ 5 1.5. Activity Evaluation Process Summary ..................................................................................... 7 1.6. Content of this Guidance ........................................................................................................ 9
2. Defense in Depth Philosophy ......................................................................................................... 10
3. Definitions and Applicability of Terms ............................................................................................ 11 3.1. Activity Evaluation................................................................................................................ 11 3.2. Licensing Basis Event Previously Evaluated in the FSAR (As Updated) ................................... 12 3.3. Design Function ................................................................................................................... 12 3.4. Change ................................................................................................................................. 13 3.5. Design Bases (Design Basis) .................................................................................................. 15 3.6. Facility as Described in the FSAR (As Updated) ..................................................................... 15 3.7. Final Safety Analysis Report (As Updated) ........................................................................ 1615 3.8. Input Parameters ................................................................................................................. 16 3.9. Malfunction of Safety-Significant SSCs.................................................................................. 17 3.10. Methods of Evaluation ......................................................................................................... 17 3.11. Departure from a Method of Evaluation Described in the FSAR (As Updated) ....................... 18 3.12. Procedures as Described in the FSAR (As Updated) .............................................................. 20 3.13. Safety Analyses .................................................................................................................... 20 3.14. Activity Screening................................................................................................................. 21 3.15. Tests or Experiments Not Described in the FSAR (As Updated) ............................................. 21
4. Implementation Guidance ............................................................................................................. 23 v

4.1. Applicability ......................................................................................................................... 23 4.2. Activity Screening................................................................................................................. 28 4.2.1.1. Screening of Changes to the Facility as Described in the USFAR ................ 32 4.2.1.2. Screening of Changes to Procedures as Described in the UFSAR ............... 34 4.2.1.3. Screening Changes to UFSAR Methods of Evaluation ................................ 35 4.3. Evaluation Process ............................................................................................................... 38 4.3.8.1. DID Examples ........................................................................................... 50 4.3.9.1. Guidance for Changing One or More Elements of a Method of Evaluation ................................................................................................ 53 4.3.9.2. Guidance for Changing from One Method of Evaluation to Another ......... 54 4.3.9.3. Examples.................................................................................................. 57 4.4. Utilizing an Activity Evaluation to Assess Compensatory Actions to Address Nonconforming or Degraded Conditions .............................................................................. 58 4.5. Disposition of Activity Evaluations .................................................................................... 5960

5. Documentation and Reporting ....................................................................................................... 61

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 List of Abbreviations AOO Anticipated Operational Occurrence ANS American Nuclear Society ANSI American National Standards Institute ASME American Society of Mechanical Engineers BDBE Beyond Design Basis Event CFR Code of Federal Regulations COL Combined construction and operating license CP Construction Permit DBA Design Basis Accident DBE Design Basis Event DBHL Design Basis Hazard Level DID Defense-in-Depth ECCS Emergency Core Cooling System EDG Emergency Diesel Generator F-C Frequency-Consequence FP Fire Protection FSAR Final Safety Analysis Report IDP Integrated Decision-Making Process IDPP Integrated Decision-Making Process Panel LBE Licensing Basis Event LMP Licensing Modernization Project LOCA Loss of Coolant Accident LWR Light water reactor NEI Nuclear Energy Institute non-LWR Non-light water reactor NRC Nuclear Regulatory Commission NSRST Non-Safety-Related with Special Treatment NST Non-Safety-Related with No Special Treatment NUMARC Nuclear Management and Resources Council NUREG Nuclear Regulatory Commission technical report designation OL Operating license PDC Principle Design Criteria PRA Probabilistic Risk Assessment RCP Reactor Coolant Pump RIPB Risk-informed and performance-based RSF Required Safety Function SAR Safety Analysis Report SER Safety Evaluation Report SG Steam Generator SR Safety-Related SGTR Steam Generator Tube Rupture SSCs Structures, Systems, and Components TEDE Total effective dose equivalent vii

UFSAR Updated Final Safety Analysis Report TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07

1. INTRODUCTION 1.1. Background 10 CFR 50.59 establishes a process and criteria for determining the regulatory threshold for when the Nuclear Regulatory Commission (NRC) must review and approve a proposed change to the facility before its implementation. 10 CFR 50.59 is not a determination of safety nor of overall acceptability. It defines the boundary between those proposed changes to the facility that can be implemented by the licensee without prior NRC approval and those that must receive NRC review and approval before implementation. Nuclear Energy Institute (NEI) document NEI 96-07 Guidelines for 10 CFR 50.59 Implementation provides guidance for applying the 10 CFR 50.59 criteria to currently operating light water reactors (LWRs). NRC endorsed the use of NEI 96-07 in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59. This guidance was developed based on the existing change control guidance in NEI 96-07, with appropriate additions and adjustments as provided herein as needed to address the risk-informed and performance-based (RIPB) licensing approach utilized in NEI 18-04 and documented per NEI 21-07.

This regulatory change guidance is predicated on the same logic as 10 CFR 50.59. Specifically, Commented [A1]: The comparison of the 50.59 requirements to the guidance in the TIRICE document has been removed from this guidance is oriented around design functions rather than potential alterations of the Updated this revision. It would seem reasonable to include this comparison Final Safety Analysis Report (UFSAR) content. Further, this guidance will contains a provision as a way of demonstrating that TIRICE addresses all of the elements analogous to 10 CFR 50.59( c) (4), commonly known as Applicability Determinations. The contained in 50.59.

aggregate effects on UFSAR content of this regulatory change process and any other applicable regulation are managed by 10 CFR 50.71(e). As explained in more detail in Section 1.3, the guidance and associated criteria in this document are intended to take the place of 10 CFR 50.59 for certain non-LWR nuclear power reactors with a safety analysis based on the methodology endorsed in RG 1.233, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors.1 and be invoked by a license condition in combination with an exemption, in whole or in part, to 10 CFR 50.59. Commented [A2]: Should this sentence be moved to be the beginning of the paragraph? It would clarify the intent.

Recommend this move and delete the first or modify the first 1.1.1. NEI 18-04 Risk-Informed Performance-Based Technology Inclusive Guidance for Non- sentence.

Light Water Reactor Licensing Basis Development NEI 18-04 Revision 12 presents a technology-inclusive, RIPB process for selection of Licensing Basis Events (LBEs); safety classification of Structures, Systems, and Components (SSCs) and associated risk-informed special treatments; and determination of Defense-in-Depth (DID) adequacy for non-LWR SSCs including, but not limited to, molten salt reactors, high-temperature gas cooled reactors, and a variety of fast reactors at all thermal power capacities.

NRC endorsed the methodology in Regulatory Guide 1.233, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis 1 NRC ADAMS Accession Number ML20091L698, June 2020.

2 NRC ADAMS Accession Number ML19241A336, August 2019.

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and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors.3 Significant attributes of the methodology that relate to the application of change control are summarized below.

  • The probabilistic risk assessment (PRA) plays a central role in the identification of LBEs, quantification of their frequencies and consequences, and evaluation of their risk significance.

AOOs, DBEs, and BDBEs are composed of event sequence families identified and evaluated in the PRA.

  • DBAs are defined using a set of deterministic rules that include the identification of Required Safety Functions (RSFs). DBAs are derived from DBEs but rely upon only Safety-Related (SR) SSCs for performance of the RSFs, and the DBAs are evaluated conservatively, with consequences compared against the same dose criteria applied to light water reactor (LWR) DBAs.
  • The remaining LBEs (AOOs, DBEs, and BDBEs) are defined and evaluated as part of the development of the PRA, using realistic assumptions and inputs and evaluation of uncertainties consistent with the non-light water reactor (non-LWR) PRA standard.
  • A systematic process for establishing the baseline system requirements is used to ensure that plant capabilities and programs are sufficient to ensure safety-significant functions meet established acceptance criteria and provide adequate DID. This includes selecting targets for Commented [A3]: Words added for completeness reliability and capability that inform the selection of special treatments to ensure safety-significant SSCs can accomplish their required functions and identifying SSCs and associated human actions that provide plant capabilities and programmatic measures to ensure adequate DID.

NEI 18-04 addresses how to establish an Licensing Modernization Project (LMP)-based safety case. That safety case becomes part of the licensing basis of the reactor when NRC issues a 10 Commented [A4]: Generic Comment: This document includes use of the terms affirmative safety case, safety case, and CFR Part 50 or Part 52 operating license for the reactor (or certifies the design under Part 52). licensing case. The NRC staff notes that NEI 18-04 and RG 1.233 Nothing described in this guidance affects the substance of that initial LMP-based safety case. do not use these terms. To avoid confusion and potential This guidance applies only to activities that take place subsequent to initial licensing, which may unforeseen consequences, applicants using the LMP process should instead continue to use the established terminology in the current involve changes that impact the licensing basis. regulatory framework, including use of safety analysis and licensing basis.

1.1.2. NEI 21-07 Technology Inclusive Guidance for Non-Light Water Reactors - Safety Analysis Report Content for Applicants Using the NEI 18-04 Methodology NEI 21-07 Revision 14 provides guidance for developing portions of the Safety Analysis Report (SAR) for reactor applicants that utilize NEI 18-04. The guidance describes eight chapters of a SAR related directly to the implementation of the NEI 18-04 methodology. The chapters do not follow the standard LWR SAR outline as provided in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition. The intent of the guidance is to help ensure completeness of information submitted to NRC while avoiding

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 unnecessary burden on the applicant and rightsizing the content of the application commensurate with the complexity of the design being reviewed.

Significant attributes of the methodology that relate to the application of change control are summarized below.

  • The document describes the LMP-based affirmative safety case, which is developed through the application of the NEI 18-04 methodology. Commented [A5]: Generic comment: Would it be more consistent and appropriate to reference RG 1.233 instead of NEI 18-
  • Applicants are expected to describe the PRA at a summary level and provide key results 04 or LMP throughout the document? See the second paragraph of related to the LMP-based affirmative safety case. Section 1.1 above to see how RG 1.233 could be referenced.

NRC is developing a regulatory guide that will address the acceptability of using NEI 21-07 to develop portions of a SAR for an applicant that has utilized NEI 18-04. NRC also plans to issue guidance for developing the remaining portions of the SAR (i.e., those portions not covered by NEI 21-07) and for other elements of a license application as part of its Advanced Reactor Content of Application Project.5 1.2. Application of this Guidance Sections 3, 4, and 5 provide change control guidance for reactors licensees following NEI 18-04 and NEI 21-07. This guidance should be authorized for use by an enabling license condition and an exemption, in whole or part, to 10 CFR 50.59. This guidance allows licensees to implement appropriate change control programs for the operation of their reactors. This guidance was developed based on the existing change control guidance in NEI 96-07, with appropriate additions and adjustments as provided herein.

It should be noted that some of the examples included in this guidance are LWR-based but still remain useful for illustrating a point. In order to put any of the examples into context for a specific design, it is necessary to explain how the change impacts the safety case for that specific design. For example, in some reactors, the Emergency Diesel Generators (EDGs) might be categorized as NST, so changes in diesel generator-start time may have no impact on the LMP safety case.

3 NRC ADAMS Accession Number ML20091L698, June 2020.

4 ADAMS Accession Number ML22060A190, February 2022.

5 Slides from the February 25, 2021, NRC Advanced Reactor Stakeholder Meeting provide information on the ARCAP project and its relationship with the TICAP project. See ML21055A541 pp.91-105.

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1.3. Purpose and Scope The purpose of this guidance is to provide a process for developing effective and consistent evaluations of proposed changes, tests, and experiments, hereafter referred to collectively as activities.

For currently licensed light water reactors, activities involving changes to plant design and operation and the conduct of new tests and experiments have the potential to affect the frequencies and consequences of accidents, create new accidents, and impact the integrity of fission product barriers. Therefore, these activities are subject to 10 CFR 50.59.

This guidance and associated criteria are intended to take the place of 10 CFR 50.59 and be invoked by a license condition in combination with an exemption, in whole or in part, to 10 CFR 50.59. Proposed changes, tests, and experiments that satisfy one or more of the criteria specified in this guidance must be reviewed and approved by NRC before implementation.

Thus, the criteria identified in this guidance provide a threshold for regulatory reviewnot the final determination of safetyfor proposed activities. Specifically, the process and criteria identified in this guidance establish the conditions under which licensees may make changes to the facility or procedures and conduct tests or experiments without prior NRC approval.

The guidance is applicable only to non-LWR licensees that implemented NEI 18-04, consistent with Regulatory Guide 1.233. The NEI 18-04 methodology is also referred to as the LMP methodology. Licensees that follow NEI 18-04 are also expected to use NEI 21-07. This guidance assumes that these two guidance documents have been used in obtaining a license from NRC.

The NEI 18-04 methodology relies on information from a PRA, and the NEI 21-07 guidance anticipates that the PRA will conform to ANSI/ASME/ANS RA-S-1.4-2021, Probabilistic Risk Assessment Standard for Advanced non-Light Water Reactor Nuclear Power Plants (referred to herein as the non-LWR PRA standard). This guidance applies to licensees that follow NEI 18-04, NEI 21-07, and the non-LWR PRA standard. Licensees that deviate from elements of NEI 18-04, NEI 21-07, or the non-LWR PRA standard must justify the application of this Commented [A6]: Shouldn't this text also point to NRC RG 1.233 for completeness?

guidance to change control. For example, an applicant would need to address its approach for its PRA.

NEI 21-07 provides guidance for developing reactor SARs for certain licensing pathways, including: 10 CFR Part 52 combined construction and operating license (COL) without reference to a design certification or an early site permit; 10 CFR Part 50 construction permit (CP) followed by an operating license (OL); and 10 CFR Part 52 design certification. This guidance does not address the design certification pathway, so its scope is limited to the 10 CFR Part 52 COL (no design certification or early site permit) and the 10 CFR Part 50 CP/OL. Commented [A7]: The NRC staff reiterates that this limitation could limit the usefulness of this document.

The objectives of this guidance are to:

  • Establish a clear understanding and process for how the criteria for making changes to the facility as described in the final safety analysis report (as updated) without prior NRC approval may be met

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07

  • Ensure that the changes that require NRC prior approval are properly identified.
  • Minimize the unnecessary burden to the regulator and operators for determining if changes require a license amendment Change control under 10 CFR 50.59 is only one of many regulations that apply to nuclear power reactors. The regulation addresses the need for prior NRC approval for certain changes to a facility licensed under 10 CFR Part 50 or 10 CFR Part 52. Other regulatory processes address areas such as operability, reportability, corrective action, and changes in the state of knowledge and are not addressed herein.

1.4. Relationship of this Guidance to Regulatory Requirements and Controls As the process for controlling a range of activities affecting equipment and procedures at a nuclear power plant, implementation of this guidance interfaces with many other regulatory requirements and controls. To optimize the use of this guidance, it should be understood in the context of the proper relationship with these other regulatory processes. These relationships are described below.

1.4.1. Relationship of this Guidance to Other Processes that Control Licensing Basis Activities This guidance focuses on the effects of proposed activities on the safety analyses contained in the UFSAR and is a cornerstone of each plants licensing basis. In addition to control of changes affecting the safety analyses, there are several other complementary processes for controlling activities that affect other aspects of the licensing basis, as described below.

  • Amendments to the operating license (including the technical specifications) are sought and obtained under 10 CFR 50.90.
  • Where changes to the facility or procedures are controlled by more specific regulations (e.g.,

quality assurance, security, and emergency preparedness program changes controlled under 10 CFR 50.54(a), (p), and (q), respectively), the more specific regulation applies.

  • Changes that require an exemption from a regulation are processed in accordance with 10 CFR 50.12.
  • Guidance for controlling changes to licensee commitments is provided by NEI 99-04, Guideline for Managing NRC Commitment Changes6.
  • Where a licensee possesses a license condition that specifically permits changes to the NRC-approved fire protection program (i.e., has received the standard fire protection license condition contained in NRC Generic Letter 86-10), subsequent changes to the fire protection program would be controlled under the license condition and not this guidance.
  • Maintenance activities, including associated temporary changes, are subject to the technical specifications and are assessed and managed in accordance with the Maintenance Rule, 10 CFR 50.65; screening and evaluation under this guidance are not required.

6 NRC Regulatory Issue Summary 00-017, Managing Regulatory Commitments Made by Power Reactor Licensees to the NRC Staff documents the NRC position that NEI 99-04 describes an acceptable way for licensees to control regulatory commitments. ADAMS Accession Number ML003741774).

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Together with the process provided in this guidance, these processes form a framework of complementary regulatory controls over the licensing basis. To optimize the effectiveness of these controls and minimize duplication and undue burden, it is important to understand the scope of each process within the regulatory framework. This guideline discusses the scope of this guidance in relation to other processes, including circumstances under which different processes (e.g., 10 CFR 50.90) should be applied to different aspects of an activity.

In addition to controlling changes to the facility and procedures described in the UFSAR, licensees may also choose to commit to control changes to other licensing basis information using the process provided in this guidance. This may be in accordance with a requirement of the license or commitment to NRC. The technical specifications bases are an example of documentation that may be outside the UFSAR, but that could be controlled via application of the process provided in this guidance.

1.4.2. Relationship of this Guidance to 10 CFR Part 50, Appendix B In NEI 21-07, quality assurance is included as a part of the Principal Design Criteria (PDC).

Prior to the operating license, 10 CFR Part 50, Appendix B, assures that the facility design and construction meet applicable requirements, codes, and standards in accordance with the safety classification SSCs. Appendix B design control provisions ensure that all changes continue to meet applicable design and quality requirements. The design and licensing bases evolve in accordance with Appendix B requirements up to the time that an operating license is received, and this guidance is not applicable until after that time. Both Appendix B and this guidance, when invoked by a license condition in combination with an exemption to 10 CFR 50.59, apply following receipt of an OL or COL.

Appendix B also addresses corrective action. The application of this guidance to compensatory actions that address degraded and nonconforming conditions is described in Section 4.4.

1.4.3. Relationship of this Guidance to the UFSAR This guidance delineates a process that identifies when a license amendment is required prior to implementing changes to the facility or procedures described in the UFSAR or tests and experiments not described in the UFSAR. As such, it is important that the UFSAR be properly maintained and updated in accordance with 10 CFR 50.71(e). Guidance for updating UFSARs to reflect activities implemented under this guidance is provided by Regulatory Guide 1.181, which endorses NEI 98-03 Revision 1. Reactors utilizing NEI 18-04 and NEI 21-07 should substitute the 10 CFR 50.59 controls described in NEI 98-03, Revision 1 with the process described in this guidance.

1.4.4. Relationship of this Guidance to 10 CFR 50.2 Design Bases This guidance controls changes to both 10 CFR 50.2 design bases and supporting design information contained in the UFSAR. In support of implementation, Section 4.3.9 provides guidance on the scope of methods of evaluation used in establishing design bases or in the safety analyses that are subject to control under Criterion (i) (see Section 4.3.9). Additional guidance

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 for identifying 10 CFR 50.2 design bases is provided in NEI 97-04, Appendix B, Guidance and Examples for Identifying 10 CFR 50.2 Design Bases.

As described in Section 3.3, RSFs represent one category of Design Basis Functions as described Commented [A8]: Added for completeness within NEI 97-04. The controls of the methodology used in establishing design bases or in the safety analyses are implemented with Section 4.3.9 of this guidance. As described in Section 2.0, the controls established in 10 CFR 50.59(c)(2)(vii)) of design basis limits for fission product barriers are implemented through other criteria, specifically Criteria (f) and (g) of Section 4.3. Commented [A9]: Added Section 4.3. Is this the correct section?

1.5. Activity Evaluation Process Summary After determining that a proposed activity is safe and effective through appropriate engineering and technical evaluations, the activity evaluation process is applied to determine if a license amendment is required prior to implementation. The activity evaluation process involves the following basic steps, as depicted in Figure 1:

  • Applicability and Screening: Determine if a proposed activity evaluation is required.
  • Evaluation: Apply the nine evaluation criteria of Section 4.3.1 through 4.3.9 (Criteria (a) -

(i)) to determine if a license amendment must be obtained from NRC.

  • Documentation and reporting: Document and report to NRC activities implemented under the activity evaluation process.

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Proposed Activity Evaluate Activity to Ensure It Is Safe and Effective Does the Activity Require a Yes Request/Obtain a License Amendment Change to the Under 10 CFR 50.90 Tech Specs?

See Section 4.1 Applicability No Determination Apply Other Regulation/Process, e.g.,

Is the Activity

  • 10 CFR Part 20 No See Is an Activity Section 4.2 No Activity Evaluation Implement the Activity Screening Required?

Yes Perform an Activity Evaluation See Section 4.3 Does the Activity Yes Request/Obtain a License Amendment Evaluation Require Prior Under 10 CFR 50.90 Process NRC Approval?

No Document Evaluation, Implement the Activity and Report to NRC per Section 4.5 Figure 1. Activity Screening and Evaluation Process

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 Later sections of this document discuss key definitions, provide guidance for determining applicability, screening, and performing activity evaluations, and present examples to illustrate the application of the process.

1.6. Content of this Guidance NRC has established requirements for nuclear plant systems, structures, and components to provide reasonable assurance of adequate protection of public health and safety. Many of these requirements, and descriptions of how they are met, are documented in the UFSAR. When followed, the process delineated in this guidance allows a licensee to determine if NRC approval is required before making changes in the facility or procedures as described in the UFSAR and to conduct tests or experiments not described in the UFSAR unless the changes require a change in the technical specifications or otherwise require prior NRC approval. In order to perform activity screenings and evaluations, an understanding of the design and licensing basis of the plant and of the specific requirements of the regulations is are necessary. Individuals performing activity screenings and evaluations should also understand the concepts discussed in this guidance.

In Section 2, the approach for determining the adequacy of DID is discussed as background for applying the guidance.

Section 3 presents definitions and a discussion of key terms used in this guidance.

This section includes guidance on the applicability requirements for the use of this guidance, the screening process for determining when an activity evaluation must be performed and the evaluation criteria for determining if prior NRC approval is required. Section 4 discusses the application of the criteria presented in Sections 4.3.1 through 4.3.9 to the process of changing the plant or procedures and the conduct of tests or experiments. Examples are provided to reinforce the guidance. Guidance is also provided on addressing degraded and nonconforming conditions and on dispositioning activity evaluations.

Section 5 provides guidance on documenting activity evaluations and reporting to NRC.

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2. DEFENSE IN DEPTH PHILOSOPHY DID,7 the use of multiple independent complementary layers of defense for protecting the public from potential harm from nuclear reactor operation, is an important part of the design, licensing, and operation of nuclear power plants. NEI 18-04 provides a technology inclusive RIPB method of implementation of the NRCs longstanding DID philosophy. It enables a systematic means for establishing DID adequacy and creating a baseline of information in the design records and the Final Safety Analysis Report (FSAR) as described in NEI 21-07 Revision 1. The baseline of information serves as the foundation of 10 CFR 50.59 change evaluations that are relevant to maintaining DID adequacy.

The DID adequacy evaluation summarized in the FSAR is structured around three elements:

Plant Capability DID, Programmatic Capability DID, and the integrated DID evaluation. The licensing baseline information for these elements is distributed across multiple chapters of an LMP SAR based on the methodology endorsed in RG 1.233. Changes to DID adequacy relevant Commented [A10]: See comment in Section 1.1.2.

to the activity evaluation process are discussed in this guidance in Section 4.3.8. This guidance utilizes the layers of defense framework as described in NEI 18-04, including the explicit considerations of individual changes in LBE performance margins, changes in SSC safety significance, changes in uncertainties, and integrated risk topics that aggregate impacts across LBEs. Thresholds for NRC review of proposed changes in Section 4.3 are adapted to the LMP terminology and the RIPB integrated decision-making methodology endorsed by RG 1.233 such that adequate DID remains assured when the attributes for DID adequacy described in NEI 18-04 Chapter 5 are confirmed for the change. These thresholds enable focused evaluations of DID to determine when a license amendment request is required.

7 A more detailed definition of DID is provided in the NRC glossary provided in Appendix A to NEI 21-07.

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07

3. DEFINITIONS AND APPLICABILITY OF TERMS The following definitions and terms are discussed in this section:

Section Definitions and Terms 3.1 Activity Evaluation 3.2 Licensing Basis Event Previously Evaluated in the FSAR (as updated) 3.3 Design Function 3.4 Change 3.5 Design Bases (Design Basis) 3.6 Final Safety Analysis Report (as updated) 3.7 Input Parameters 3.8 Final Safety Analysis Report (as updated) 3.9 Malfunction of an SSC Important to Safety 3.10 Methods of Evaluation 3.11 Departure from a Method of Evaluation Described in the FSAR (as updated) 3.12 Procedures as Described in the FSAR (as updated) 3.13 Safety Analyses 3.14 Activity Screening 3.15 Tests or Experiments Not Described in the FSAR (as updated) 3.1. Activity Evaluation Definition An activity evaluation is the documented evaluation against the nine criteria in Sections 4.3.1 through 4.3.9 to determine if a proposed change, test, or experiment requires prior NRC approval via license amendment under 10 CFR 50.90.

Discussion It is important to establish common terminology for use relative to the activity evaluation process. The definitions of Screening and Activity Evaluation and Screening are intended to clearly distinguish between the process and documentation of licensee screenings and the further evaluation that may be required of proposed activities against the nine criteria in Sections 4.3.1 through 4.3.9. Section 4.3 provides guidance for performing activity evaluations. The screening process is discussed in Section 4.2.

The phrase change made under this guidance (or equivalent) refers to changes subject to this guidance (see Section 4.1) that either screened out of the activity evaluation process or did not require prior NRC approval based on the results of an activity evaluation. Similarly, the phrases this guidance applies [to an activity] or [an activity] is subject to an activity screening and evaluation mean that screening and, if necessary, evaluation are required for the activity. The 11

evaluation process includes screening, evaluation, documentation, and reporting to NRC of activities subject to the application of this guidance. Commented [A11]: Paragraph is a bit hard to follow. It could benefit from a bulleted list or making separate paragraphs.

3.2. Licensing Basis Event Previously Evaluated in the FSAR (As Updated)

Definition ThetermLBEpreviouslyevaluatedintheFSAR(asupdated)meansadesignbasisaccidentorotherlicensingbasisevent(LBE)describedintheUFSARincludingLBEs,suchasthosein Chapter 3 of the UFSAR per NEI 21-07, transients (i.e., bounded LBEs) and events the facility is required to withstand such as floods, fires, earthquakes, other external hazards.

Discussion The term "licensing basis event" refers to the AOOs, DBEs, BDBEs, and DBAs that are analyzed to demonstrate that the facility can be operated without undue risk to the health and safety of the public. For a reactor with an LMP-based affirmative safety case, LBEs are defined as the AOOs, DBEs, BDBEs, and DBAs, which, as noted above, are documented in SAR Chapter 3 per NEI 21-07. Note that, although fire is an event for which a plant is required to cope and is described in the UFSAR, changes to the fire protection program are, for most licensees, governed by requirements other than this guidance, as discussed in Section 4.1.5.

Other transient and events the facility is required to withstand include those that define design basis hazard levels (DBHLs) as addressed in SAR Chapters 2 and 6 per NEI 21-07.

3.3. Design Function Definition Design functions that are related to the LMP-based safety case include the safety significant functions performed by SSCs classified as SR and NSRST as described in Sections 6 and 7 of the UFSAR prepared according to NEI 21-07. As defined in NEI 96-07, with respect to 10 CFR 50.59 screening, design functions are UFSAR-described design bases functions and other SSC functions described in the UFSAR that support or impact design bases functions. The term design bases functions in the context of an LMP based safety case cover SSCs that perform Required Safety Functions, risk significant functions, or functions needed for adequate Defense-in-Depth. Therefore, the scope of 10 CFR 50.2 design bases functions described in this guidance Commented [A12]: It appears that it is inconsistent with that under Discussion below which states:

document is a subset of design functions.

For the purpose of evaluating changes to a reactor with an LMP-Design bases functions are functions performed by SSCs that are (1) required by, or otherwise based affirmative safety case, design bases functions correspond to RSFs per NEI 21-07 SAR Section 5.2.

necessary to comply with applicable regulations, license conditions, orders, or technical specifications, or (2) credited in the safety analyses to meet NRC requirements. Implicitly Section 3.5 also states that LMP RSFs represent Design Basis included within the meaning of design function are the conditions under which intended Functions as described in Section 3.3.

functions are required to be performed, such as equipment response times, process conditions, Should this be edited to remove the word bases?

equipment qualification (See Regulatory Guide 1.186 and NEI 97-04 Appendix B as appropriate).

Commented [A13]: Replaced non-safety-related with NSRST to be consistent with the two paragraphs above. Non-UFSAR description of design functions may identify what SSCs are intended to do, when and safety-related SSCs do not perform safety significant functions.

how design functions are to be performed, and under what conditions. Design functions may be Commented [A14]: This doesn't seem correct or consistent performed by safety-related SSCs or non-safety-relatedNSRST SSCs and include functions that, if not performed, with NEI 18-04. A malfunctioning SSC could (and in many cases would) be a part of an evaluated event sequence, and not the wcould initiate a transient (i.e., a bounded LBE) or accident that the plant is required to withstand. initiator of a transient or accident.

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 The phrase required by, or otherwise necessary to comply with, regulations is directed toward the design-specific PDCs as described within its FSAR/application. Design functions that accommodate a specific PDC may be outside the scope of the SR and NSRST SSCs described in Chapter 6 and 7 of a NEI 21-07 SAR.

Discussion For the purpose of evaluating changes to a reactor with an LMP-based affirmative safety casewith a safety analysis based on the methodology endorsed in RG 1.233, design bases functions correspond to RSFs per NEI 21-07 SAR Section 5.2. Design functions are Commented [A15]: This is too narrowly defined. As described above in Section 3.3, the term design bases functions in the considered to be composed of the RSFs, risk-significant functions per NEI 21-07 SAR Section context of an LMP based safety case cover SSCs that perform 5.5.1, and other safety functions required for adequate DID per NEI 21-07 SAR Section 5.5.2. Required Safety Functions, risk significant functions, or functions needed for adequate Defense-in-Depth.

As used above, credited in the safety analyses means that if the SSC were not to perform its design bases function in the manner described, the assumed initial conditions, mitigative actions, or other information in the analyses would no longer be within the range evaluated (i.e., the analysis results would be called into question). The phrase support or impact design bases functions refers both to those SSCs needed to support design bases functions (cooling, power, environmental control, etc.) and to SSCs whose operation or malfunction could adversely affect the performance of design bases functions (for instance, control systems and physical arrangements).

3.4. Change Definition Change means a modification or addition to, or removal from, the facility or procedures that affects: (1) a design function, (2) method of performing or controlling the function, or (3) an evaluation that demonstrates that intended functions will be accomplished.

Discussion Additions to and removals from the facility or procedures can adversely impact the performance of SSCs and the bases for the acceptability of their design and operation. Thus, the definition of change includes modifications of an existing provision (e.g., SSC design requirement, analysis method or parameter), additions or removals (physical removals, abandonment, or non-reliance on a system to meet a requirement) to the facility or procedures.

The definitions of change (see Section 3.4), facility (see Section 3.6), and procedures (see Section 3.12) make clear that this guidance applies to changes to underlying analytical bases for the facility design and operation as well as for changes to SSCs and procedures. Thus, this activity evaluation process should be applied to a change being made to an evaluation for demonstrating adequacy of the facility even if no physical change to the facility is involved.

Further discussion of the terms in this definition is provided as follows:

Method of performing or controlling a function means how a design function is accomplished as credited in the safety analyses, including specific operator actions, procedural step, or sequence, or whether a specific function is to be initiated by manual versus automatic means.

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For example, substituting a manual actuation for automatic would constitute a change to the method of performing or controlling the function.

Evaluation that demonstrates that intended functions will be accomplished means the method(s) used to perform the evaluation (as discussed in Section 3.10). Example: a thermodynamic calculation that demonstrates an auxiliary cooling system has sufficient heat removal capacity for responding to a DBA.

See Section 4.2 for a discussion of adverse effects and their role in the overall screening process.

Changes to Special Treatment The term special treatment is used in a manner consistent with NRC regulations and Nuclear Energy Institute (NEI) guidelines in the implementation of 10 CFR 50.69. In Regulatory Guide 1.201, the following definition of special treatment is provided:

special treatment refers to those requirements that provide increased assurance beyond normal industrial practices that structures, systems, and components (SSCs) perform their design-basis functions.

A distinction is made between special treatment as applied to SR SSCs and alternative special Commented [A16]: And NSRST?

treatment afforded by 10 CFR 50.69. Alternative treatment requirements are differentiated from special treatment requirements in the use of reasonable confidence versus reasonable assurance, which is a general conclusion in initial plant licensing. Commented [A17]: What is the difference between reasonable confidence and reasonable assurance? It would be useful to explain the difference in the text.

Special treatments are applied to safety significant (SR and NSRST SSCs) to ensure they accomplish their design functions with adequate reliability and capability. SSCs must be capable of performing their design functions in the expected accident conditions in which they will be called upon to perform. Therefore, any non-administrative change to a special treatment identified in the UFSAR (as updated) that has an adverse effect on a design function constitutes a change to be evaluated in accordance with Section 4.3.

Temporary Changes Temporary changes to the facility or procedures, such as jumpering terminals, lifting leads, placing temporary lead shielding on pipes and equipment, removal of barriers and use of temporary blocks, bypasses, scaffolding and supports, are made to facilitate a range of plant activities and are subject to this activity evaluation process as follows:

  • An activity evaluation should be applied to temporary changes proposed as compensatory actions to address degraded or nonconforming conditions as discussed in Section 4.4.
  • Other temporary changes to the facility or procedures that are not associated with maintenance are subject to an activity evaluation in the same manner as permanent changes, to determine if prior NRC approval is required. Screening and, as necessary, evaluation of such temporary changes may be considered as part of the screening/evaluation of the proposed permanent change.

Risk impacts of temporary changes associated with maintenance activities (i.e., temporary alterations) should be assessed and managed in accordance with 10 CFR 50.65(a)(4) and

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 associated guidance, as discussed in Section 4.1.2. Performing an activity evaluation for such activities is not required provided that temporary alterations are not in effect longer than 90 days at power, and affected SSCs are restored to their normal, as-designed condition at the conclusion of the maintenance activity.

3.5. Design Bases (Design Basis)

Definition Design bases means that information which identifies the specific functions to be performed by a structure, system, or component of a facility 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 calculations and/or experiments) of the effects of a design basis accident for which a structure, system, or component must meet its functional goals.

Discussion LMP RSFs represent one category of Design Basis Functions as described in Section 3.3. Commented [A18]: Same comment as above - This is too narrowly defined. As described above in Section 3.3, the term Guidance and examples for identifying 10 CFR 50.2 design bases are provided in Appendix B of design bases functions in the context of an LMP based safety case NEI 97-04, Design Bases Program Guidelines, Revision 1, November 2000. cover SSCs that perform Required Safety Functions, risk significant functions, or functions needed for adequate Defense-in-Depth.

3.6. Facility as Described in the FSAR (As Updated)

Definition Facility as described in the final safety analysis report (as updated) means:

  • The SSCs that are described in the FSAR (as updated),
  • The design and performance requirements for such SSCs described in the FSAR (as updated), and
  • The evaluations or methods of evaluation included in the FSAR (as updated) for such SSCs, which demonstrate that their intended function(s) will be accomplished.

Discussion The scope of information that is the focus of activity screenings and evaluations performed using this guidance, is the information presented in the original FSAR to satisfy the requirements of 10 CFR 50.34(b), as updated per the requirements of 10 CFR 50.71(e) and as supplemented pursuant to 10 CFR 54.21(d). The definition of facility as described in the FSAR (as updated) Commented [A19]: This is the first time 10 CFR 54 is mentioned. If TIRICE is to be applicable during the license renewal follows from the requirement of 10 CFR 50.34(b) that the FSAR (and by extension, the UFSAR) period, it should be mentioned in Section 1.3, Purpose and Scope.

contains a description and analysis of the SSCs of the facility, with emphasis upon performance requirements, the bases, with technical justification therefore, upon which such requirements have been established, and the evaluations required to show that safety functions will be accomplished.

An activity screening of facility changes is discussed in Section 4.2.1.1.

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3.7. Final Safety Analysis Report (As Updated)

Definition Final Safety Analysis Report (as updated) means the Final Safety Analysis Report (or Final Hazards Summary Report) submitted in accordance with 10 CFR 50.34, as amended and supplemented, and as updated per the requirements of 10 CFR 50.71(e) or 10 CFR 50.71(f), as Commented [A20]: The previous page has as supplemented pursuant to 10 CFR 54.21(d) for renewed licenses. Should it be applicable. For Part 52 COLs, the content of the Final Safety Analysis Report should be in added here? For all instances where 50.71(e) appears in this accordance with the requirements specified in 10 CFR 52.77, Contents of applications; general guidance, they should be evaluated for consistency.

information. Commented [A21]: Should this instead be 52.79, Contents of applications; technical information in final safety analysis report,?

Discussion As used throughout this guidance, UFSAR is synonymous with FSAR (as updated). The scope of the UFSAR includes its text, tables, and diagrams, as well as supplemental information explicitly incorporated by reference. References that are merely listed in the UFSAR and documents that are not explicitly incorporated by reference are not considered part of the UFSAR and therefore are not subject to control under this guidance and associated license condition.

Use of this guidance for activity screening and evaluation is not required for UFSAR information that is subject to other specific change control regulations. For example, licensee quality assurance programs, emergency plans and security plans are controlled by 10 CFR 50.54(a), (p) and (q), respectively.

The FSAR (as updated), for purposes of this guidance, also includes UFSAR update pages approved by the licensee for incorporation in the UFSAR since the last required update was submitted per 10 CFR 50.71(e). The intent of this requirement is to ensure that decisions about proposed activities are made with the most complete and accurate information available.

Pending UFSAR revisions may be relevant to a future activity that involves that part of the UFSAR. Therefore, pending UFSAR revisions to reflect completed activities that have received final approval for incorporation in the next required update should be considered as part of the UFSAR for purposes of activity screenings and evaluations, as appropriate. Appropriate configuration management mechanisms should be in place to identify and assess interactions between concurrent changes affecting the same SSCs or the same portion of the UFSAR.

Guidance on the required content of UFSAR updates is provided in Regulatory Guide 1.181 and NEI 98-03, Revision 1, Guidelines for Updating FSARs, June 1999.

3.8. Input Parameters Definition Input parameters are those values derived directly from the physical characteristics of SSC or processes in the plant, including flow rates, temperatures, pressures, dimensions or measurements (e.g., volume, weight, and size), and system response times.

Discussion The principal intent of this definition is to distinguish methods of evaluation from evaluation input parameters. Changes to methods of evaluation described in the UFSAR (see Section 3.10)

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 are evaluated under Criterion (i) in Section 4.3.9, whereas changes to input parameters described in the FSAR are considered changes to the facility that would be evaluated under the other eight criteria in Sections 4.3.1 through 4.3.8, but not under Criterion (i). Commented [A22]: The NRC staff have a concern with this approach. As discussed in comments below, a change to an input parameter/assumption in an analysis would likely not be flagged If a methodology permits the licensee to establish the value of an input parameter on the basis of under the other criterion. NEI 96-07 states that a change to plant-specific considerations, then that value is an input to the methodology, not part of the inputs/assumptions involves a change to a method of evaluation.

methodology. On the other hand, an input parameter is considered to be an element of the methodology if:

  • The method of evaluation includes a methodology describing how to select the value of an input parameter to yield adequately conservative results. However, if a licensee opts to use a value more conservative than that required by the selection method, reduction in that conservatism should be evaluated as an input parameter change, not a change in methodology.
  • The development or approval of a methodology was predicated on the degree of conservatism in a particular input parameter or set of input parameters. In other words, if certain elements of a methodology or model were accepted on the basis of the conservatism of a selected input value, then that input value is considered an element of the methodology.

Examples illustrating the treatment of input parameters are provided in Section 4.2.1.3.

Section 4.3.9 provides guidance and examples to describe the specific elements of evaluation methodology that would require evaluation under Criterion (i) and to clearly distinguish these from specific types of input parameters that are controlled by the other eight criteria of Sections 4.3.1 through 4.3.8.

3.9. Malfunction of Safety-Significant SSCs Definition Malfunction of SR or NSRST SSCs (i.e., SSCs important to safety) means the failure of SSCs to perform their intended design functions or capabilities described in the UFSAR.

Discussion For the purpose of evaluating changes to a reactor with an a safety analysis based on the methodology endorsed in RG 1.233LMP-based affirmative safety case, SSCs important to safety are defined to be the population of SSCs that are either SR or NSRST SSCs, as defined by NEI 18-04. This population of SSCs is also referred to as the safety-significant SSCs.

Guidance and examples for applying this definition are provided in Section 4.3.

3.10. Methods of Evaluation Definition Methods of evaluation means the calculational framework used for evaluating behavior or response of the facility within its design bases.

17

Discussion Examples of methods of evaluation are presented below. Changes to such methods of evaluation require evaluation under 10 CFR 50.59(c)(2)(viii) only for evaluations used either in UFSAR safety analyses or in establishing the design bases, and only if the methods are described, outlined, or summarized in the UFSAR. Commented [A23]: Wouldnt the results of the PRA be described in the FSAR? A clarification may be necessary.

Methodology changes that are subject to an activity evaluation include changes to elements of existing methods described in the UFSAR and to changes that involve replacement of existing methods of evaluation with alternative methodologies.

Examples

  • Fuel performance models
  • Radionuclide inventory models
  • Mechanistic source term models
  • Vendor-specific thermal design procedure
  • Dose conversion factors and assumed source term(s) (e.g., ICRP factors) Commented [A24]: Global comment: Acronym uses. DNBR, ASME etc. Please spell out acronyms with initial uses.
  • Data correlations (e.g., DNBR correlations) Commented [A25]: This sticks out as an LWR requirement.
  • Means of data reduction (e.g., ASME III and Appendix G methods for evaluating reactor Would heat transfer correlations be more appropriate?

vessel embrittlement specimens)

  • Physical constants or coefficients (e.g., heat transfer coefficients)
  • Mathematical models (e.g., decay heat models)
  • Specific limitations of a computer program
  • Specified factors to account for uncertainty in measurements or data
  • Statistical treatment of results Commented [A26]: Would analysis be a better term to use?

Methods of evaluation described in the UFSAR subject to Section 4.3.9 Criterion (i) are:

  • Methods of evaluation used in UFSAR safety analyses, including analyses typically presented in UFSAR Chapters 2 and/or 3 per NEI 21-07, to demonstrate that consequences of DBAs do not exceed 10 CFR 100 or 10 CFR 50.34, dose limits Commented [A27]: Part 100 and 50.34 are regulatory dose criteria not limits.
  • Methods of evaluation used in supporting UFSAR analyses that demonstrate intended design Commented [A28]: What about the integrated risk criteria functions will be accomplished under design basis conditions that the plant is required to contained in NEI 18-04? Another bullet should be added that withstand, including natural phenomena, environmental conditions, and dynamic effects as requires the integrated risk criteria be included in the evaluation.

reflected in the definition and evaluation of the Design Basis Hazard Levels (DBHLs) Commented [A29]: Do these examples include aircraft impact and LOLA? If not, should they?

3.11. Departure from a Method of Evaluation Described in the FSAR (As Updated) Commented [A30]: Section 3.11 seems to cover the same topic as Section 3.10. Should they be combined?

Definition Departure from a method of evaluation described in the FSAR (as updated) means (i) changing any of the elements of the method described in the FSAR (as updated) unless the results of the analysis are conservative or essentially the same; or (ii) changing from a method described in the

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 FSAR to another method unless that method has been approved by NRC for the intended application.

Discussion The definition of departure provides licensees with flexibility to make changes in methods of evaluation that are conservative or that are not important with respect to demonstrating that SSCs can perform their intended design functions. See also the definition and discussion of Methods of Evaluation in Section 3.9. Guidance for evaluating changes in methods of Commented [A31]: Should this be Section 3.10?

evaluation under Criterion (i) is provided in Section 4.3.9.

Conservative vs. Nonconservative Evaluation Results Gaining margin by revising an element of a method of evaluation is considered to be a nonconservative change and thus a departure from a method of evaluation for purposes of this guidance. Such departures require prior NRC approval of the revised method. In other words, analytical results obtained by changing any element of a method are conservative relative to the previous results, if they are closer to design bases limits or safety analyses limits (e.g.,

applicable acceptance guidelines). For example, a change in an element of a method of evaluation that changes the result of a containment peak pressure analysis from 45 psig to 48 psig (with a design basis limit of 50 psig) would be considered a conservative change for purposes of an activity evaluation under Criterion (i) provided in Section 4.3.9. This is because results closer to limiting values are considered conservative in the sense that the new analysis result provides less margin to applicable limits for making future physical or procedure changes without a license amendment.

If use of a modified method of evaluation resulted in a change in calculated containment peak pressure from 45 psig to 40 psig, this would be nonconservative. This is because the change would result in more margin being available (to the design basis limit of 50 psig) for a licensee to make more significant future changes to the physical plant or procedures.

Essentially the Same Licensees may change one or more elements of a method of evaluation such that results move in the nonconservative direction without prior NRC approval, provided the results are essentially the same as the previous result. Results are essentially the same if they are within the margin of error for the type of analysis being performed. Variation in results due to routine analysis sensitivities or calculational differences (e.g., rounding errors and use of different computational platforms) would typically be within the analysis margin of error and thus considered essentially the same.

Approved by NRC for the Intended Application Rather than make a minor change to an existing method of evaluation, a licensee may also adopt completely new methodology without prior NRC approval provided the new method is approved by NRC for the intended application. A new method is approved by NRC for the intended application if it is approved for the type of analysis being conducted and the licensee satisfies applicable terms and conditions for its use. Specific guidance for making this determination is provided in Section 4.3.9.

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3.12. Procedures as Described in the FSAR (As Updated)

Definition Procedures as described in the final safety analysis report (as updated) means those procedures that contain information described in the FSAR (as updated) such as how structures, systems, and components are operated and controlled (including assumed operator actions and response times).

Discussion The scope of information that is the focus of the application of the activity screening and evaluation process is the information presented in the original FSAR to satisfy the requirements of 10 CFR 50.34(b), as updated per the requirements of 10 CFR 50.71(e) and as supplemented pursuant to 10 CFR 54.21(d). Commented [A32]: See previous comment regarding 10 CFR 54.

For purposes of this guidance, procedures are not limited to plant procedures specifically identified in the UFSAR (e.g., operating and emergency procedures). Procedures include UFSAR descriptions of how actions related to system operation are to be performed and controls over the performance of design functions. This includes UFSAR descriptions of operator action sequencing or response times, certain descriptions (text or figure) of SSC operation and operating modes, operational and radiological controls, and similar information. If changes to these activities or controls are made, such changes are considered changes to procedures described in the UFSAR, and the changes are subject to the activity screening and evaluation process.

Even if described in the UFSAR, procedures that do not contain information on how SSCs are operated or controlled do not meet the definition of procedures as described in the UFSAR and are not subject to the activity screening and evaluation process. Sections 4.1.2 and 4.1.4 identify examples of procedures that are not subject to an activity screening and evaluation.

Activity screening of procedure changes is discussed in Section 4.2.1.2.

3.13. Safety Analyses Definition Safety analyses are analyses performed pursuant to NRC requirements to demonstrate the integrity of the reactor coolant pressure boundary, the capability to shut down the reactor and maintain it in a safe shutdown condition, or the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines in 10 CFR 50.34(a)(1) or 10 CFR 100.11. Safety analyses are required to be Commented [A33]: This definition is LWR-centric (based on 10 CFR 50.2) and/or may not capture all elements of a safety analysis.

presented in the UFSAR per 10 CFR 50.34(b) and 10 CFR 50.71(e) and include, but are not Should it be revised?

limited to, the DBA analyses typically presented in Chapter 2 and/or 3 of the UFSAR as defined in NEI 21-07.

Discussion Safety analyses are those analyses or evaluations that demonstrate that acceptance criteria for the facilitys capability to withstand or respond to design basis accidents are met. Safety analyses include DBA analyses and the deterministic analyses that are done to ensure the SR SSCs are

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 able to perform their RSFs following a DBHL. DBA analyses are presented in Chapters 2 and/or 3 of the UFSAR clearly fall within the meaning of safety analyses as defined above. Also, within the meaning of this definition for purposes of this guidance are:

  • Supporting UFSAR analyses that demonstrate that SSC design functions will be accomplished as credited in the accident analyses
  • UFSAR analyses of events that the facility is required to withstand, such as turbine missiles, fires, floods, earthquakes Note that, although fire is an event that a plant is required to withstand and for which it has been analyzed accordingly in the UFSAR (by reference to the Fire Hazards Analysis for some licensees), changes to the fire protection program and associated analyses are (for most licensees) governed by licensee requirements other than this activity evaluation process, as discussed in Section 4.1.5.

3.14. Activity Screening Definition Screening is the process for determining whether a proposed activity requires an activity evaluation to be performed.

Discussion Screening is that part of the process that determines whether an activity evaluation is required prior to implementing a proposed activity.

The definitions of change, facility as described, procedures as described and test or experiment not described constitute criteria for the screening process. Activities that do not meet these criteria are said to screen out from further review under an activity evaluation, i.e.,

they may be implemented without an activity evaluation.

Engineering and technical information concerning a proposed activity may be used along with other information as the basis for determining if the activity screens out or requires an activity evaluation.

Further discussion and guidance on screening are provided in Section 4.2.

3.15. Tests or Experiments Not Described in the FSAR (As Updated)

Definition Tests or experiments not described in the final safety analysis report (as updated) means any activity where any structure, system, or component is utilized or controlled in a manner which is either:

  • Outside the reference bounds of the design bases as described in the UFSAR, or
  • Inconsistent with the analyses or descriptions in the UFSAR.

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Discussion Activity screening and evaluation is applied to tests or experiments not described in the UFSAR.

The intent of the definition is to ensure that tests or experiments that put the facility in a situation that has not previously been evaluated (e.g., unanalyzed system alignments) or that could affect the capability of SSCs to perform their intended design functions (e.g., high flow rates, high temperatures) are evaluated before they are conducted to determine if prior NRC approval is required.

Maintenance-related testing is assessed and managed under 10 CFR 50.65(a)(4), as discussed in Section 4.1.2. Activity screening of tests and experiments unrelated to maintenance is discussed in Section 4.2.2. Examples of tests unrelated to maintenance and thus subject to activity screening and evaluation include (1) most core physics testing, (2) room heat-up testing to validate a design/analysis input, and (3) testing to help determine which of two redesign alternatives to pursue.

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07

4. IMPLEMENTATION GUIDANCE Licensees may perform applicability determination and adverse impact screening activities to determine if activity evaluations are required as described in Sections 4.1 and 4.2, or equivalent manner.

4.1. Applicability This guidance is applicable only to non-LWR reactor licensees with a safety analysis based on the methodology endorsed in RG 1.233that implemented NEI 18-04, consistent with Regulatory Guide 1.233. Licensees that follow NEI 18-04 are also expected to conform to NEI 21-07, Commented [A34]: Global comment: The applicability of this guidance is stated differently throughout this document. For which provides guidance for developing reactor SARs for certain licensing pathways: 10 CFR example, under Section 1.3 it says:

Part 52 COL without reference to a design certification or an early site permit; 10 CFR Part 50 CP followed by an OL; and 10 CFR Part 52 design certification. This guidance does not address This guidance applies to licensees that follow NEI 18-04, NEI 21-07, and the non-LWR PRA standard. Licensees that deviate from the design certification pathway, so its applicability is limited to the 10 CFR Part 52 COL (no elements of NEI 18-04, NEI 21-07, or the non-LWR PRA standard design certification or early site permit) and the 10 CFR Part 50 CP/OL. must justify the application of this guidance to change control.

A consistent statement on applicability is suggested throughout this This guidance applies to each holder of a license authorizing operation of a production or guidance.

utilization facilitycommercial nuclear reactor, that has a license condition that invokes the use of this guidance and an approved exemption, in whole or in part, from 10 CFR 50.59. This guidance also covers the holder of a license authorizing operation of a nuclear power reactor that has submitted a certification of permanent cessation of operations required under 10 CFR 50.82(a)(1) or a reactor licensee whose license has been amended to allow possession but not operation of the facility.

4.1.1. Applicability to Licensee Activities The use of this guidance is required and applicable to tests or experiments not described in the UFSAR and to changes to the facility or procedures as described in the UFSAR, including changes made in response to new requirements or generic communications, except as noted below:

  • Proposed activities that require a change to the technical specifications must be made via the license amendment process, 10 CFR 50.90. Aspects of proposed activities that are not directly related to the required technical specification change are subject to activity screening and evaluation.
  • To reduce duplication of effort, changes to the facility or procedures that are controlled by other more specific requirements and criteria established by regulation are specifically excluded from the scope of activities that are required to be evaluated using this guidance.

For example, 10 CFR 50.54, specifies criteria and reporting requirements for changing Commented [A35]: See previous comment on 10 CFR 54.

quality assurance, physical security, and emergency plans.

In addition to 50.90 and 50.54(a), (p) and (q), the following include change control requirements that meet the intent of applicable regulations that establish more specific criteria for accomplishing such changes and may take precedence over the use of this guidance for control of specific changes:

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  • 10 CFR 50.65 (Maintenance Rule) (See additional discussion in Section 4.1.2.)
  • Standard Fire Protection license condition (if applicable) (See additional discussion in Section 4.1.5.)

Activities controlled and implemented under other regulations may require related information in the UFSAR to be updated. To the extent the UFSAR changes are directly related to the activity implemented via another regulation, applying activity screening and evaluation is not required.

UFSAR changes should be identified to NRC as part of the required UFSAR update, per 10 CFR 50.71(e). However, there may be certain activities for which a licensee would need to apply both the requirements of this screening and evaluation process and that of another regulation. For example, a modification to a facility involves additional components and substantial piping reconfigurations as well as changes to protection system setpoints. The protection system setpoints are contained in the facility technical specifications. Thus, a license amendment to revise the technical specifications under 10 CFR 50.90 is required to implement the new system setpoints. The activity screening and evaluation should be applied to the balance of the modification, including impacts on required operator actions.

4.1.2. Maintenance Activities Maintenance activities are activities that restore SSCs to their as-designed condition, including activities that implement approved design changes. Maintenance activities are not subject to this screening and evaluation process but are subject to the provisions of 10 CFR 50.65(a)(4) as well as technical specifications.

Maintenance activities include troubleshooting, calibration, refurbishment, maintenance-related testing, identical replacements, housekeeping and similar activities that do not permanently alter the design, performance requirements, operation, or control of SSCs. Maintenance activities also include temporary alterations to the facility or procedures that directly relate to and are necessary to support the maintenance. Examples of temporary alterations that support maintenance include jumpering terminals, lifting leads, placing temporary lead shielding on pipes and equipment, removal of barriers, and use of temporary blocks, bypasses, scaffolding, and supports.

Licensees should ensure operability in accordance with the technical specifications and should assess and manage the risk impact of maintenance activities per 10 CFR 50.65(a)(4) and NUMARC 93-01, Industry Guidelines for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants.8 8

Regulatory Guide 1.182, issued June 1, 2000, endorses the industry guidance on 10 CFR 50.65(a)(4) provided in Section 11 of NUMARC 93-01, Revision 3, August 2000.

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 In addition to assessments required by 10 CFR 50.65(a)(4), this screening and evaluation process should also be applied in the following cases:

  • A temporary alteration in support of the maintenance is expected to be in effect during at-power operations for more than 90 days. In this case, this screening and evaluation process would be applied to the temporary alteration prior to implementation in the same manner as a permanent change. Commented [A37]: What is the basis for 90 days?
  • The plant is not restored to its original condition upon completion of the maintenance activity (e.g., if SSCs are removed, the design, function or operation is altered, or if temporary alteration in support of the maintenance is not removed). In this case, this screening and evaluation process would be applied to the permanent change to the plant.

Installation and post-modification testing of approved facility changes are indistinguishable, in terms of their risk impact on the plant, from maintenance activities that restore SSCs to their as-designed condition. As such, installation and testing of approved facility changes are maintenance activities that must be assessed and managed in accordance with 10 CFR 50.65(a)(4). This screening and evaluation process will address the effect, following implementation, of proposed facility changes to determine if prior NRC approval is required; the risk impact of actually implementing the change will be assessed and managed per 10 CFR 50.65(a)(4).

If a temporary alteration necessary to install a facility change is expected to be in effect longer than 90 days at power, the required screening and evaluation review of the temporary alteration may be performed as part of the screening and evaluation process review for the facility change.

This screening and evaluation process should be applied to temporary changes proposed as compensatory actions for degraded or nonconforming conditions, as discussed in Section 4.4.

Control of Maintenance Procedures Changes to procedures for performing maintenance are made in accordance with applicable 10 CFR Part 50, Appendix B criteria and licensee procedures. Licensee processes should ensure that changes to plant configurations called for by procedures are consistent with the technical specifications. This screening and evaluation process does not apply to such changes because, Commented [A38]: Should this sentence also include a confirmation that plant configuration changes are consistent with like the maintenance activities themselves, changes to procedures for performing maintenance do the UFSAR, in addition to the technical specifications?

not permanently alter the design, performance requirements, operation, or control of SSCs.

Certain maintenance procedures, including those for technical specification required surveillance and inspection, may contain important information concerning SSC design, performance, operation, or control. Examples include acceptance criteria for valve stroke times or other SSC function, torque values, and types of materials (gaskets, elastomers, lubricants, etc.). Licensee design and/or configuration control processes should ensure that this screening and evaluation process is applied to changes in such information and that maintenance procedure changes do not inadvertently alter the design, performance requirements, operation, or control of SSCs.

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If a change to a maintenance procedure affects information in the UFSAR (e.g., a specific test or maintenance frequency), the affected information should be updated in accordance with 10 CFR 50.71(e).

4.1.3. UFSAR Modifications Modifications to the UFSAR that are not the result of activities performed under this screening and evaluation process are not subject to control under this guidance. Such modifications include reformatting and simplification of UFSAR information and removal of obsolete or redundant information and excessive detail.

Similarly, this screening and evaluation process need not be applied to the following types of activities:

  • Editorial changes to the UFSAR (including referenced procedures, topical reports, etc.)
  • Clarifications to improve reader understanding
  • Correction of inconsistencies within the UFSAR (e.g., between sections)
  • Minor corrections to drawings, e.g., correcting mislabeled valves
  • Similar changes to UFSAR information that do not change the meaning or substance of information presented 4.1.4. Changes to Procedures Governing the Conduct of Operations Even if described in the UFSAR, changes to managerial and administrative procedures governing the conduct of facility operations are controlled under 10 CFR Part 50, Appendix B, programs and are not subject to control under this screening and evaluation process. These include, but are not limited to, procedures in the following areas:
  • Operations and work process procedures such as control of equipment status (tag outs)
  • Shift staffing and personnel qualifications Commented [A39]: Changes to important human actions should be supported by an HFE and be included within this
  • Changes to position titles screening. Per SRP 18.0: "An HFE review is conducted if such a modification affects the role of personnel or the tasks they
  • Administrative controls for creating or modifying procedures perform, the sequence of actions, the timing, or the overall workload, and is potentially significant to plant safety."
  • Training programs
  • On-site/off-site safety review committees
  • Plant modification process
  • Calculation process Example 1 The UFSAR states that the shift supervisor will authorize all radioactive liquid releases. This is an administrative requirement on the conduct of facility operations. Thus, assigning this function to another individual would not be subject to this screening and evaluation process but would be done in accordance with 10 CFR Part 50, Appendix B criteria and licensee procedures.

The licensee would be required to reflect the change in the next required update of the UFSAR, per 10 CFR 50.71(e).

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 4.1.5. Changes to Approved Fire Protection Programs Most nuclear power plant licenses contain a section on fire protection (FP). Originally, these fire protection license conditions varied widely in scope and content. These variations created problems for licensees and for NRC inspectors in identifying the operative and enforceable fire protection requirements at each facility.

To resolve these problems, NRC promulgated guidance in Generic Letter 86-10, Implementation of Fire Protection Requirements, for licensees to:

  • Amend the operating license to substitute a standard fire protection license condition for the previous license condition(s) regarding fire protection.

Under the standard fire protection license condition, licensees may:

Make changes to their approved FP programs without prior NRC approval provided that the changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire, and Alter specific features of the approved program provided such changes do not otherwise involve a change to the license or technical specifications or require an exemption.

Adoption of the standard fire protection license condition provided a more consistent approach to evaluating changes to the facility, including those associated with the fire protection program.

Originally, changes to the FP program under the FP license condition were also subject to this screening and evaluation process; however, this created confusion as to which regulatory requirement governed FP program changes.

It is important to note that when applicable regulations establish more specific criteria for controlling certain changes, this screening and evaluation process does not also apply.

Consistent with this intent, the standard fire protection license condition establishes specific criteria for control of FP changes and falls outside the scope of Section 4.3. Thus, applying this screening and evaluation process to fire protection program changes is not required.

Changes to the fire protection program should be evaluated for impacts on other design functions, and this screening and evaluation process should be applied to the non-fire protection related effects of the change, if any.

Consistent with current practice, determinations made under the standard fire protection license condition should be based on a written evaluation that remains available for NRC review for the life of the plant. These written evaluations should provide the basis for the licensees conclusion that changes to the fire protection program do not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. An evaluation performed in accordance with the license condition should include an assessment of the impact of the change on the existing fire hazards analysis for the area, as is current practice. The assessment should address the effects on 27

combustible loading and distribution and should consider whether circuits or components, including associated circuits, for a train of equipment needed for safe shutdown could be affected, or whether a new element could be introduced into the area.

Under the standard license condition, approved fire protection program documents (e.g., fire hazards analysis) are incorporated in the UFSAR, and as such, changes to this information are subject to 10 CFR 50.71(e) reporting requirements.

4.1.6. New Information New Information is routinely acquired as a natural part of the design, operation, maintenance of nuclear facilities. New information for new reactor designs is expected to be identified as a result of the compilation of the knowledge from operating experience, experiments, and testing.

It is recognized that new information could impact the safety case and would be subject to other Commented [A40]: "licensing basis" would be a better term here, and would better align with the Background and Scope regulatory controls (e.g., 10 CFR Part 50 Appendix B, 10 CFR 50.150 and 10 CFR 50.155). In sections of the document.

that regard, new information may lead to changes that are subject to an activity screening and potentially evaluation (e.g., a change to a method of evaluation due to an evolution of the understanding of a particular physical phenomenon, a plant modification to regain margin loss due to the availability of new information, and a plant modification to take advantage of margin gained by the availability of new information). However, new information, in and of itself, is not subject to review using this guidance.

The most commonly encountered of these regulations would be corrective action. If corrective action is taken to resolve this condition, then this guidance would not be applicable.

4.2. Activity Screening Once it has been determined that this activity screening and evaluation process is applicable to a proposed activity, screening is performed to determine if the activity should be evaluated against the nine evaluation criteria in Sections 4.3.1 through 4.3.9.

Engineering, design, and other technical information concerning the activity and affected SSCs should be used to assess whether the activity is a test or experiment not described in the UFSAR or a modification, addition, or removal (i.e., change) that affects:

  • A design function of an SSC
  • A method of performing or controlling the design function, or
  • An evaluation for demonstrating that intended design functions will be accomplished.

Sections 4.2.1 and 4.2.2 provide guidance and examples for determining whether an activity is (1) a change to the facility or procedures as described in the UFSAR or (2) a test or experiment not described in the UFSAR. If an activity is determined to be neither, then it screens out and may be implemented without further evaluation under this activity screening and evaluation.

Activities that are screened out from further evaluation under this activity screening and evaluation should be documented as discussed in Section 4.2.3.

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 Each element of a proposed activity must be screened except in instances where linking elements of an activity is appropriate, in which case the linked elements can be considered together. A test for linking elements of proposed changes is interdependence.

It is appropriate for discrete elements to be considered together if (1) they are interdependent as in the case where a modification to a system or component necessitates additional changes to other systems or procedures; (2) they are performed collectively to address a design or operational issue; or (3) one or more of the elements are planned to ensure that the overall change preserves DID adequacy (see Criterion (h) in Section 4.3 of this document). To state it another way, it is allowable to include an additional element or elements in the screening and evaluation of a proposed change if that additional element or elements are being performed to address what might otherwise be an undesirable result for Criterion (h).

An example of the first criterion is a pump upgrade modification may also necessitate a change to a support system, such as cooling water. An example of the second criterion is reinforcing structural members in support of the installation of a heavier component. An example of the third criterion is a change in special treatment that is linked to preserving a single safety function identified in order to maintain adequate DID.

If concurrent changes are being made that are not linked, each must be screened separately and independently of each other. If, however, there are changes to the DID baseline that are made to offset any adverse impacts of a change to DID adequacy through applying the change process in NEI 18-04 Sections 5.9.6 and 5.9.7, such concurrent changes are considered linked.

Activities that screen out may nonetheless require UFSAR information to be updated. Licensees should provide updated UFSAR information to NRC in accordance with 10 CFR 50.71(e).

Specific guidance for applying this activity screening and evaluation process to temporary changes proposed as compensatory actions for degraded or nonconforming conditions is provided in Section 4.4.

4.2.1. Is the Activity a Change to the Facility or Procedures as Described in the USFAR?

To determine whether a proposed activity affects a design function, method of performing or controlling a design function or an evaluation that demonstrates that design functions will be accomplished, a thorough understanding of the proposed activity is essential. For the LMP-based affirmative safety case, design functions should be documented in Chapter 5 of the SAR per NEI 21-07a Required Safety Function in Section 5.2, a risk-significant function in Section 5.5.1, and a safety function required for adequate DID in Section 5.5.2. A given activity may have both direct and indirect effects that the screening review must consider. The following questions illustrate a range of effects that may stem from a proposed activity:

  • Does the activity decrease the reliability or capability of an SSC design function, including Commented [A41]: Added for completeness either functions whose failure would initiate a transient/ LBE or functions that are relied upon for mitigation?
  • Does the activity reduce defense-in-depth?

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  • Does the activity add or delete an automatic or manual design function of the SSC?
  • Does the activity convert a feature that was automatic to manual or vice versa?
  • Does the activity introduce an unwanted or previously unreviewed system or materials interaction?
  • Does the activity adversely affect the ability or response time to perform required actions, e.g., alter equipment access or add steps necessary for performing tasks?
  • Does the activity degrade the ability of the SSC to withstand design basis hazards (e.g.,

seismic) or environmental qualification of the SSC?

  • Does the activity adversely affect other units at a multiple unit site?
  • Does the activity affect a method of evaluation used in establishing the design bases or in the safety analyses? Commented [A42]: Same comment as earlier, if changes to evaluation inputs or assumptions are not considered a change to a
  • For activities affecting SSCs, procedures, or methods of evaluation that are not described in method of evaluation, then it does not appear that these other the UFSAR, does the change have an indirect effect on electrical distribution, structural criteria will pick this up.

integrity, environmental conditions, or other UFSAR-described design functions?

Per the definition of change discussed in Section 3.4, this guidance is applicable to additions as well as to changes to and removals from the facility or procedures. Additions should be screened for their effects on the existing facility and procedures as described in the UFSAR and, if required, this activity evaluation process should be performed. NEI 98-03 provides guidance for determining whether additions to the facility and procedures should be reflected in the UFSAR per 10 CFR 50.71(e).

Consistent with historical practice, changes affecting SSCs or functions not described in the UFSAR must be screened for their effects (so-called indirect effects) on UFSAR-described design functions. This activity screening and evaluation is required when such changes adversely affect a UFSAR-described design function, as described below.

Screening for Adverse Effects An activity screening is required for changes that adversely affect design functions, methods used to perform or control design functions, or evaluations that demonstrate that intended design functions will be accomplished (i.e., adverse changes). Changes that have none of these effects, or have positive effects, may be screened out because only adverse changes have the potential to increase the likelihood of malfunctions, increase consequences, create new LBEs or otherwise meet the evaluation criteria delineated in Sections 4.3.1 through 4.3.9.

Per the definition of design function, SSCs may have preventive, as well as mitigative, design functions. Adverse changes to either must be screened in. Thus, a change that decreases the reliability of a function whose failure could initiate an LBE would be considered to adversely affect a design function and would screen in. In this regard, changes that would relax the manner in which code requirements are met for certain SSCs should be screened for adverse effects on design function. Similarly, changes that would introduce a new type of LBE or malfunction would screen in. This reflects an overlap between the technical/engineering (safety) review of the change and an activity screening and evaluation performed in accordance with this guidance.

This overlap reflects that these considerations are important to both the safety and regulatory reviews.

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 If a change has both positive and adverse effects, the change should be screened in. The evaluation should focus on the adverse effects. Commented [A43]: Is a tradeoff allowed (i.e., what if the positive effects out weigh the adverse effects)?

The screening process is not concerned with the magnitude of adverse effects that are identified. Commented [A44]: If a tradeoff between positive and adverse effects is allowed, this sentence would need to be deleted.

Any change that adversely affects a UFSAR-described design function, method of performing or controlling design functions, or evaluation that demonstrates that intended design functions will be accomplished is screened in. The magnitude of the adverse effect (e.g., is the minimal increase standard met?) is the focus of the evaluation process.

Screening determinations are made based on the engineering/technical information supporting the change. The screening focus on design functions, etc., ensures the essential distinction between activity screenings, and evaluations, which focus on whether changes meet any of the nine criteria in Sections 4.3.1 through 4.3.9. Technical/engineering information (e.g., design evaluations) that demonstrates changes have no adverse effect on UFSAR-described design functions, methods of performing or controlling design functions, or evaluations that demonstrate that intended design functions will be accomplished may be used as basis for screening out the change. If the effect of a change is such that existing safety analyses would no longer be bounding and therefore UFSAR safety analyses must be re-run to demonstrate that all required safety functions and design requirements are met, the change is considered to be adverse and must be screened in. The revised safety analyses may be used in support of the required evaluation of such changes.

Changes that entail update of safety analyses to reflect improved performance, capacity, timing, etc., resulting from a change (beneficial effects on design functions) are not considered adverse and need not be screened in, even though the change calls for safety analyses to be updated. For example, a change that improves the closure time of main control room isolation dampers reduces the calculated dose to operators, and UFSAR dose consequence analyses are to be updated as a result. In this case, the dose analyses are being revised to reflect the lower dose for the main control room, not to demonstrate that applicable dose limits continue to be met. A change that would adversely affect the design function of the dampers (isolation of the main control room) and increase the existing calculated dose to operators would be considered adverse and would screen in. In this case, the dose analyses must be re-run to ensure that applicable dose limits continue to be met. The revised analyses would be used in support of the change evaluation to determine if the increase exceeds the minimal standard and requires prior NRC approval.

To further illustrate the distinction between activity screening and evaluation, consider the example of a change to a diesel generator-starting relay that delays the diesel start time from 10 seconds to 12 seconds. The UFSAR-described design function credited in the Emergency Core Cooling System (ECCS) analyses is for the diesel to start within 12 seconds. This change would screen out because it is apparent that the change will not adversely affect the diesel generator design function credited in the ECCS analyses (ECCS analyses remain valid).

However, a change that would delay the diesels start time to 13 seconds would screen in because the change adversely effects the design function (to start in 12 seconds). Such a change would screen in even if technical/engineering information supporting the change includes revised 31

safety analyses that demonstrate all required safety functions supported by the diesel (core heat removal, containment isolation, containment cooling, etc.) are satisfied and that applicable dose limits continue to be met. While this change may be acceptable with respect to performance of required safety functions and meeting design requirements, the analyses necessary to demonstrate acceptability are beyond the scope/intent of activity screening reviews. Thus, an activity evaluation would be required. The revised safety analyses would be used in support of the activity evaluation to determine whether any of the criteria are met such that prior NRC approval is required for the change. Additional specific guidance for identifying adverse effects due to a procedure or methodology change is provided in Sections 4.2.1.2 and 4.2.1.3, respectively.

4.2.1.1. Screening of Changes to the Facility as Described in the USFAR Screening to determine that a change evaluation is required is straightforward when a change adversely affects an SSC design function, method of performing or controlling a design function, or evaluation that demonstrates intended design functions will be accomplished as described in the UFSAR. Commented [A45]: This should be clarified. Same comment as previous, does this apply to changes to inputs to evaluations that NEI 96-07 considers changes to methods of evaluation?

SSCs that are relied upon to carry out design functions are documented in the NEI 21-07 SAR in Section 5.4 (SR SSCs) and Section 5.5 (NSRST SSCs). However, as addressed in Section 4.2.1.1, changes to other SSCs (i.e., NST SSCs) should be considered for potential adverse effects on any SR or NSRST SSC design function, method of performing or controlling the design function, or an evaluation demonstrating that the intended design functions will be accomplished.

In accordance with NEI 18-04, reliability and capability targets are documented in NEI 21-07 SAR Sections 6.2 and 7.1 for SR and NSRST SSCs, respectively. If a proposed change to the facility results in a safety-significant SSC being unable to meet its reliability or capability target, then the change would screen in, and a full evaluation in accordance with Section 4.3 would be required.

A facility also contains many SSCs not described in the UFSAR. These can be components, subcomponents of larger components or even entire systems. Changes affecting SSCs that are not explicitly described in the UFSAR can have the potential to adversely affect SSC design functions that are described and thus may require an activity change evaluation. In such cases, the approach for determining whether a change involves a change to the facility as described in the UFSAR is to consider the larger, UFSAR-described SSC of which the SSC being modified is a part. If for the larger SSC, the change adversely affects a UFSAR-described design function, method of performing or controlling the design function, or an evaluation demonstrating that intended design functions will be accomplished, then an activity evaluation is required.

Another important consideration is that a change to NST SSCs not described in the UFSAR can indirectly affect the capability of SSCs to perform their UFSAR-described design function(s).

For example, increasing the heat load on a non-safety-related heat exchanger could compromise the cooling systems ability to cool safety-related equipment.

Seismic qualification, missile protection, flooding protection, fire protection, environmental qualification, high energy line break and masonry block walls are some of the areas where

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 changes to NST SSCs, whether or not described in the UFSAR, can affect the UFSAR-described design function of safety-significant SSCs through indirect or secondary effects.

Equivalent replacement is a type of change to the facility that does not alter the design functions of SSCs. Licensee equivalence assessments, e.g., consideration of performance/operating characteristics and other factors, may thus form the basis for screening determinations that no activity change evaluation is required.

As discussed in Section 4.2.1, only proposed changes to SSCs that would, based on supporting engineering and technical information, have adverse effects on design functions require an activity evaluation. Changes that have positive or no effect on design functions may generally be screened out.

The following examples illustrate the screening process as applied to proposed facility changes.

Example 1 A licensee proposes to replace a relay in the overspeed trip circuit of an emergency diesel generator with a nonequivalent relay. The relay is not described in the UFSAR, but the design functions of the overspeed trip circuit and the emergency diesel generator are. Based on engineering/technical information supporting the change, the licensee determines if replacing the relay would adversely affect the design function of either the overspeed trip circuit or EDG. If the licensee concludes that the change would not adversely affect the UFSAR-described design function of the circuit or EDG, then this determination would form the basis for screening out the change, and an activity change evaluation would not be required. This example demonstrates where the functional impact of a given change is properly assessed even though the SSC being changed is beneath the level detail in the UFSAR description.

Example 2 A licensee proposes a nonequivalent change to the operator on one of the safety injection accumulator isolation valves. The UFSAR describes that these isolation valves are open with their circuit breakers open during normal operation. These are motor operated, safety-related valves required for pressure boundary integrity and to remain open so that flow to the reactor coolant system will occur during a Loss of Coolant Accident (LOCA) as reactor coolant system pressure drops below ~600 psi. They are remotely closed during a normal shutdown so as to not inject when not required. Technical/engineering work supporting this change ensures that the replacement operator is capable of performing the functions of the existing operator and will not adversely affect the connected Class 1E bus or diesel. This change would screen out because (1) the valve operator does not perform, support, or impact the UFSAR-described design function (to ensure pressure boundary integrity and remain open when required) that supports safety injection performance credited in the safety analyses, and (2) the change does not adversely affect other SSC design functions (e.g., of the Class 1E bus).

If the proposed change was to configure the valve as a normally closed valve that automatically opens on loss of reactor coolant system pressure, an activity change evaluation would be required because the change would adversely affect the reliability of the safety injection function as credited in the safety analyses.

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Example 3 A licensee proposes to replace a globe valve with a ball valve in a vent/drain application to reduce the propensity of this valve to leak. The UFSAR-described design function of this valve is to maintain the integrity of the system boundary when closed. The vent/drain function of the valve does not relate to design functions credited in the safety analyses, and the licensee has determined that a ball valve is adequate to support the vent/drain function and is superior to the globe valve in terms of its isolation function. Thus, the proposed change affects the design of the existing vent/drain valvenot the design function (pressure boundary integrity) that supports system performance credited in the safety analysesand an activity change evaluation/reporting is not required. The screening determination should be documented, and the UFSAR should be updated per 10 CFR 50.71(e) to reflect the change.

Example 4 The bolts for retaining a rupture disk are being replaced with bolts of a different material and fewer threads, but equivalent load capacity and strength, such that the rupture disk will still relieve at the same pressure as before the change. Because the replacement bolts are equivalent to the original bolts, the design function of the rupture disk (to relieve at a specified pressure) is unaffected, and this activity may be screened out as an equivalent change.

Example 5 A waste processing system includes a trap designed to collect radionuclides. An inadvertent release of the traps contents has been evaluated as a BDBE. The service life of the trap has been evaluated and allows extending the replacement interval by a factor of two. This results in the postulated radioactive release increasing by the same factor due to a larger, maximum source term. The change has a potential to adversely affect the design function since the safety analyses assumptions may no longer be conservative. Therefore, the proposed activity screens in for further evaluation.

4.2.1.2. Screening of Changes to Procedures as Described in the UFSAR Changes are screened in (i.e., require a change evaluation) if they adversely affect how SSC design functions are performed or controlled (including changes to UFSAR-described procedures, assumed operator actions and response times). Proposed changes that are determined to have positive or no effect on how SSC design functions are performed or controlled may be screened out. Procedures should be screened in only if they affect design Commented [A46]: Adversely affect?

functions (see Sections 3.3 and 4.1.1). Required operator actions should be addressed in the SAR documentation of the associated SSCs, provided in NEI 21-07 SAR Chapter 6 (SR SSCs) and NEI 21-07 SAR Chapter 7 (NSRST SSCs).

For purposes of activity screening, changes that fundamentally alter (replace) the existing means of performing or controlling design functions should be conservatively treated as adverse and screened in. Such changes include replacement of automatic action by manual action (or vice versa), changes to the man-machine interface, changing a valve from locked closed to administratively closed and similar changes.

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 The following examples illustrate the activity screening process as applied to proposed changes affecting how SSC design functions are performed or controlled.

Example 1 If the UFSAR description of the reactor start-up procedure contains eight fundamental sequences, the licensee's decision to eliminate one of the sequences would screen in. On the other hand, if the licensee consolidated the eight fundamental sequences and did not affect the method of controlling or performing reactor start-up, the change would screen out.

Example 2 The UFSAR states that a particular flow path is isolated by a locked closed valve when not in use. A procedure change would remove the lock from this valve such that it becomes a normally closed valve. In this case, the design function is to remain closed, and the method of performing the design function has fundamentally changed from locked closed to administratively closed.

Thus, this change would screen in and require a change evaluation to be performed. Commented [A47]: This may be a topic for further discussion between industry and NRC. It is not clear how the NEI 18-04 process addresses the topic of human error, including the likelihood Example 3 of an error of commission (opening this unlocked closed valve, for Operations proposes to revise its procedures to change from 8-hour shifts to 12-hour shifts. This example). So, how would this change get evaluated by TIRICE?

change results in mid-shift rounds being conducted every 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> as opposed to every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The NRC staff agrees that it should be screened in, but this The UFSAR describes high energy line breaks including mitigation criteria. Operator action to document may want to discuss how it gets evaluated.

detect and terminate the line break is described in the UFSAR, which specifically states that 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is assumed for the pipe break to go undetected before it would be identified during operator mid-shift rounds. The change from 4- to 6-hour rounds is a change to a procedure as described in the UFSAR that adversely affects the timing of operator actions credited in the safety analyses for limiting the effects of high energy line breaks. Therefore, this change screens in, and an activity change evaluation is required.

4.2.1.3. Screening Changes to UFSAR Methods of Evaluation As discussed in Section 3.10, methods of evaluation included in the UFSAR to demonstrate that intended SSC design functions will be accomplished are considered part of the facility as described in the UFSAR. Thus, use of new or revised methods of evaluation (as defined in Section 3.10) is considered to be a change that is controlled by this guidance and needs to be considered as part of this screening step. Adverse changes to elements of a method of evaluation Commented [A48]: What is an adverse change to an element of a method of evaluation? Do you mean non-conservative?

included in the UFSAR, or use of an alternative method, must be evaluated under Criterion (i) of Section 4.3.9 to determine if prior NRC approval is required. Changes to methods of evaluation (only) do not require evaluation against the first eight criteria. Methods of evaluation associated with DBAs should be addressed in NEI 21-07 SAR Sections 2.2 (Source Term), 2.3 (DBA Analytical Methods), and 3.6 (Design Basis Accidents). Adverse changes to DBA methods utilized to demonstrate that design limits are met for SSCs that provide functions that support the retention of radioactive material would screen in. Methods of evaluation associated solely with the remaining LBEs (AOOs, DBEs, and BDBEs) should be addressed in the PRA and are not intended to be evaluated using this guidance. Methods of evaluation not associated with DBAs may be addressed in NEI 21-07 SAR Section 2.4 (Other Methodologies and Analyses) or in other parts of the SAR not covered by NEI 21-07 guidance.

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Changes to methods of evaluation not included in the UFSAR or to methodologies included in the UFSAR that are not used in the safety analyses or to establish design bases may be screened out. If a method of evaluation is used in the PRA and is also included in the UFSAR for a DBA analysis, then it would screen in for the other UFSAR analysis.

Methods of evaluation that may be identified in references listed at the end of UFSAR sections or chapters are not subject to activity screening and evaluation in accordance with this guidance unless the UFSAR states they were used for specific analyses used in establishing the design bases or in the safety analyses.

Changes to methods of evaluation included in the UFSAR are considered adverse and require evaluation using this guidance if the changes are outside the constraints and limitations associated with use of the method, e.g., identified in a topical report and/or Safety Evaluation Report (SER). If the changes are within constraints and limitations associated with use of the method, the change is not considered adverse and may be screened out.

Proposed use of an alternative method is considered an adverse change that must be evaluated under Section 4.3.9 Criterion (i).

The following examples illustrate the screening of changes to methods of evaluation.

Example 1 The UFSAR identifies the name of the computer code used for performing containment performance analyses, with no further discussion of the methods employed within the code for performing those analyses. Changes to the computer code may be screened out provided that the changes are within the constraints and limitations identified in the associated topical report and SER. A change that goes beyond restrictions on the use of the method would be considered adverse and evaluated under Section 4.3.9 Criterion (i) to determine if prior NRC approval is required.

Example 2 The UFSAR describes the methods used for atmospheric heat transfer and containment pressure response calculations contained within the CONTEMPT computer code. The code is also used for developing long-term temperature profiles (post-recirculation phase of LOCA) for environmental qualification through modeling of the residual heat removal system. Neither this application of the code nor the analysis method is discussed in the UFSAR. A revision to CONTEMPT to incorporate more dynamic modeling of the residual heat removal system transfer of heat to the ultimate heat sink would screen out because this application of the code is not described in the UFSAR as being used in the safety analyses or to establish design bases.

Changes to CONTEMPT that affect the atmospheric heat transfer or containment pressure predictions may not screen out (because the UFSAR describes this application in the safety analyses) and may require an activity evaluation.

Example 3 The steamline break mass and energy release calculations were originally performed at a power level of 105% of the nominal power (plus uncertainties) in order to allow margin for a future power up-rate. The utility later decided that it would not pursue the power up-rate and wished to

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 use the margin to address other equipment qualification issues. The steamline break mass and energy release calculations were reanalyzed, using the same methodology, at 100% power (plus uncertainties). This change would screen out as a methodology change because the proposed activity involved a change to an input parameter (% power) and not a methodology change. This change should be screened per Section 4.2.1.1 to determine if it constitutes a change to the facility as described in the UFSAR that requires evaluation under Section 4.3.9 Criterion (i). Commented [A49]: This is confusing. Section 4.3.9 (i) addresses changes to methods of evaluation. The NRC staff doesnt see how this could be a change to the facility. Under NEI 96-07, this Example 4 would be screened in under a change to method of evaluation.

The LOCA mass and energy release calculations were originally performed at a power level of 105% of the nominal power, plus uncertainties. Some of the assumptions in the analysis were identified as nonconservative, but NRC concluded in the associated SER that the overall analysis was conservative because of the use of the higher initial power. The utility later decided that it would not pursue the power up-rate and wished to use the margin to address other equipment qualification issues. The LOCA break mass and energy release calculations were reanalyzed, using the same methodology, at 100% power (plus uncertainties). This change would not screen out because the proposed activity involved a change to an input parameter that was integral to NRC approval of the methodology.

Example 5 Due to fuel management changes, core physics parameters change for a particular reload cycle.

The topical report and associated SER that describe how the core physics parameters are to be calculated explicitly allow use of either 2-D or 3-D modeling for the analysis. A change to add or remove discretionary conservatism via use of 3-D methods instead of 2-D methods or vice-versa would screen out because the change is within the terms and conditions of the SER.

4.2.2. Is the Activity a Test or Experiment Not Described in the UFSAR?

As discussed in Section 3.15, tests or experiments not described in the UFSAR are activities where an SSC is utilized or controlled in a manner that is outside the reference bounds of the design for that SSC or inconsistent with analyses or description in the UFSAR.

As discussed in Section 4.1.2, testing associated with maintenance is assessed and managed under 10 CFR 50.65(a)(4) and is not subject to activity screening and evaluation in accordance with this guidance.

Tests and experiments that are described in the UFSAR may be screened out at this step. Tests and experiments that are not described in the UFSAR may be screened out provided the test or experiment is bounded by tests and experiments that are described. Similarly, tests and experiments not described in the UFSAR may be screened out provided that affected SSCs will be appropriately isolated from the facility. Tests or experiments described in the SAR would likely be located in NEI 21-07 SAR Chapter 2, NEI 21-07 SAR Section 6.3 (SR SSCs), and NEI 21-07 SAR Section 7.2 (NSRST SSCs).

Examples of tests that would screen in at this step (assuming they were not associated with maintenance or described in the UFSAR) would be:

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Operation with fuel demonstration assemblies Commented [A50]: Suggest adding more examples and bulletizing Examples of tests that would screen out would be:

Steam generator moisture carryover tests (provided such testing is described in the UFSAR)

Balance-of-plant heat balance test Information gathering that is nonintrusive to the operation or design function of the associated SSC Commented [A51]: Suggest bulletizing 4.2.3. Screening Documentation The following record-keeping requirements apply to activity evaluations performed for activities that screened in, not to screening records for activities that screened out. However, documentation should be maintained in accordance with plant procedures of screenings that conclude a proposed activity may be screened out (i.e., that an activity evaluation was not required). The basis for the conclusion should be documented to a degree commensurate with the safety significance of the change. For changes, the documentation should include the basis for determining that there would be no adverse effect on design functions, etc. Typically, the Commented [A52]: etc. leaves a lot to interpretation by the inspector who peruses such documentation. Suggest removing it or screening documentation is retained as part of the change package. This documentation does not changing to an example.

constitute the record of changes required by this guidance, and thus is not subject to associated documentation and reporting requirements delineated in Section 5. Screening records need not be retained for activities for which an evaluation was performed or for activities that were never implemented.

4.3. Evaluation Process Once it has been determined that a given activity requires an activity evaluation, the written evaluation must address the applicable criteria of Sections 4.3.1 through 4.3.9. These nine criteria are used to evaluate the effects of proposed activities on LBEs and malfunctions previously evaluated in the UFSAR and their potential to cause LBEs or malfunctions whose effects are not bounded by previous analyses.

Criteria (a) through (h) are applicable to activities other than changes in methods of evaluation.

Criterion (i) is applicable to changes in methods of evaluation. Each activity must be evaluated against each applicable criterion. If any of the criteria are met, the licensee must apply for and obtain a license amendment per 10 CFR 50.90 before implementing the activity. The evaluation against each criterion should be appropriately documented as discussed in Section 4.5.

Sections 4.3.1 through 4.3.9 provide guidance and examples for evaluating proposed activities against the nine change evaluation criteria.

Each element of a proposed activity must undergo an activity screening and evaluation, except in instances where linking elements of an activity is appropriate, in which case the linked elements can be evaluated together. A test for linking elements of proposed changes is interdependence.

It is appropriate for discrete elements to be evaluated together if (1) they are interdependent as in the case where a modification to a system or component necessitates additional changes to other

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 systems or procedures; or (2) they are performed collectively to address a design or operational issue; or (3) one or more of the elements are planned to ensure that the overall change does not result in a change to the defense-in-depth adequacy determination. For example, a pump upgrade modification may also necessitate a change to a support system, such as cooling water.

Another example are discrete elements linked to preserving a single RSF feature which may have been identified as necessary for adequate DID.

If concurrent changes are being made that are not linked, each must be evaluated separately and independently of each other.

The effects of a proposed activity being evaluated should be assessed against each of the evaluation criteria separately. Evaluations should consider the effects of the proposed activity on operator actions.

Specific guidance for applying activity screening and evaluation to temporary changes proposed as compensatory actions for degraded or nonconforming conditions is provided in Section 4.4.

Information on LBE frequency and consequence is developed in the PRA and is provided in the SAR for AOOs, DBEs, and BDBEs. DBAs are defined using a set of deterministic rules that involve the selection of SR SSCs in the performance of RSFs and their consequences are evaluated conservatively; the DBAs have no associated frequency. For the purpose of evaluating LBEs against the change evaluation criteria delineated in Section 4.3, AOOs, DBEs, and BDBEs are addressed in terms of frequency, consequence, and risk, whereas DBA consequences are addressed deterministically based on consequences.

In the LMP context, special treatments refer to those treatments beyond the selection of commercial grade equipment necessary to achieve the reliability and capability targets for SSCs in the performance of safety significant functions. Special treatments for safety significant SSCs are documented in the UFSAR Sections 6.2 and 7.1 for SR and NSRST SSCs, respectively according to the guidance in NEI 21-07. The evaluation criteria presented below include criteria for changes that impact the definition, frequency, and consequences of LBEs and the likelihood and consequences of malfunctions of safety significant SSCs. Any changes to special treatments documented in the UFSAR are evaluated on the basis of the impacts on these metrics as illustrated in the examples presented with the criteria. Changes to special treatments that do not impact these metrics do not impact the LMP-based safety case and do not require NRC prior approval.

The nine criteria to be considered during the evaluation process are listed below and described in more detail in the Sections 4.3.1 through 4.3.9. Commented [A53]: Suggest a tenth criterion be added to address any change that would cause a BDBE to move into the DBE range or a DBE to move into the AOO range Criterion (a)Result in a change to the frequency and/or consequences of one or more AOOs, DBEs, or BDBEs documented in the final safety analysis report (as updated) in a manner that would exceed the NEI 18-04 Frequency-Consequence Target Commented [A54]: Shouldnt the test be margin to exceedance rather than exceedance of the target?

Criterion (b)Change an AOO, DBE or BDBE from non-risk significant to risk significant Commented [A55]: Could this say NEI 18-4 F-C Target or alternative design goals used in FSAR? This could capture case according to NEI 18-04 LBE risk significance criteria. where 1 Rem used for siting/EPZ size.

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Criterion (c)Result in more than a minimal increase in the consequences of a Design Basis Accident or any other event with which the plant is designed to cope described in the final safety analysis report (as updated). Commented [A56]: This seems too broad. This would imply that a more than minimal increase in a low consequence AOO would require NRC approval?

Criterion (d)Result in identifying one or more AOO, DBE, or BDBE that is (i) not previously evaluated in the UFSAR and (ii) classified as risk significant according to NEI 18-04 LBE risk significance criteria.

Criterion (e)Result in a change to any of the NEI 18-04 cumulative risk metrics that exceeds the cumulative risk targets in Section 3.3.5 of NEI 18-04.

Criterion (f)Result in an increase in the frequency and/or consequences of a malfunction of any safety-significant SSC that would change the classification of the SSC from non-risk significant to risk-significant.

Criterion (g)Result in a change in Safety-Related SSC classification.

Criterion (h)Result in a change to the performance of a safety-significant SSC that would adversely change the defense-in-depth adequacy determination.

Criterion (i)Result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses.

4.3.1. Criterion (a)

(a) Result in a change to the frequency and/or consequences of one or more AOOs, DBEs, or BDBEs documented in the final safety analysis report (as updated) in a manner that would exceed the NEI 18-04 Frequency-Consequence Target.

In evaluating this question, the first step is to identify the LBEs that have been evaluated in the UFSAR that are affected by the proposed activity. Then a determination should be made as to whether the change increases the risk such that the AOO, DBE, or BDBE exceeds the Frequency-Consequence (F-C) target (see NEI 18-04 Section 3.2.2, Task 7a). The evaluation of Criterion (a) may be performed by using the PRA to evaluate the effect on the frequency and consequences of the event consistent with NEI 18-04. Commented [A57]: Since NRC does not review PRAs, this evaluations results are only constrained by the F-C curve? Further discussion of this evaluation would be helpful.

The following examples illustrate the implementation of this criterion.

Example 1 Commented [A58]: This text indicates that the frequency of a previously identified DBE would increase to a higher frequency DBE A high-temperature gas-cooled reactor plant proposes to raise the administrative, procedural based on the relaxed wall thinning criteria.

limit related to the steam generator nominal wall thickness plugging criteria. This change will However, the figure seems to depict a change in event sequence allow tubes with more significant defects to remain in service where previously they would have frequency from BDBE to DBE. If this were the case, then a new DBA been required to be plugged. One of the likely effects of this change is an increased probability would be introduced based on this new DBE.

of a steam generator tube rupture (SGTR) DBE. A bounding PRA analysis was performed by Please clarify.

raising the SGTR initiating event frequency. The analysis concluded that allowing tubes to remain in service with more significant defects did not result in changes to SGTR DBE consequences that exceeded the NEI 18-04 F-C Target. Therefore, Criterion (a) does not require prior NRC approval for the change (see Figure 2).

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 Figure 2. Overall Plant Performance Change in Example 1 Example 2 The state emergency management agency requests that the nuclear operator provide for additional communications beyond what is in the existing emergency plan. This communicator function would need to be performed by a control room operator. Management requests plant staff evaluate the proposed change to assist in decision-making. The potential impact to the emergency plan will be evaluated under 10 CFR 50.54(q). The potential impact to the affirmative safety case screens into the TIRICE process. The PRA staff evaluates the SGTR accident sequence, and the result is increased consequences, as shown below. The performance Commented [A59]: The figure seems to depict this as a BDBE rather than an "accident". Should this be changed to "BDBE" of the additional duties raises the human error probability for performing steam generator (SG) sequence?

cooldown in response to the SGTR event, which is necessary to limit primary-to-secondary leakage and prevent SG overfill. Although the consequences of the proposed activity increase, they do not exceed the F-C curve and would not exceed Criterion (a). Therefore, Criterion (a) Commented [A60]: If this example event sequence was a DBA rather than the BDBE depicted, then this change would likely screen in based on the "more than minimal increase in consequences" scope of Criterion (c) below, correct?

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does not require prior NRC approval for the change (see Figure 3).

Figure 3. Overall Plant Performance Change in Example 2 4.3.2. Criterion (b)

(b) Change an AOO, DBE or BDBE from non-risk significant to risk significant according to NEI 18-04 LBE risk significance criteria.

The criterion is satisfied if an existing AOO, DBE, or BDBE changes its risk classification from non-risk-significant to risk-significant (see NEI 18-04 Section 3.2.2, Task 7c). NEI 18-04 Figure 3-4 provides a graphical representation of the risk significant region and the F-C target.

The evaluation of Criterion (b) may be performed by using the PRA to evaluate the effect on risk significance and the F-C target consistent with NEI 18-04. This criterion is not applicable to DBAs because risk significance applies only to AOOs, DBEs, and BDBEs The following examples illustrates the implementation of this criterion.

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 Example 1 A plant utilizes an air-operated valve to perform a function that has been identified as part of establishing adequate DID. The design utilizes one solenoid valve to port air into the actuator to open the air-operated valve, which is a fail-open valve. The current configuration has caused plant transients due to spuriously opening at power. A design change is being implemented to add a second solenoid valve in series using a redundant signal. The proposed configuration was modeled in the PRA for the purposes of evaluating whether prior NRC approval is needed. The addition of the second solenoid valve would reduce the probability that the air-operated valve would spuriously open but would raise the probability of a failure to open on demand. However, the overall risk impact did not change any LBE from non-risk significant to risk significant in accordance with the NEI 18-04 risk significance criteria. Therefore, Criterion (b) does not require prior NRC approval for the change.

Example 2 A pump is classified as NSRST because one of its PRA safety functions exceeds the SSC risk significance criteria in NEI 18-04. As a result of a PRA update which incorporates new evidence from plant specific service experience, it is determined that the pump is more reliable than assessed in the PRA used to establish the LMP safety case documented in the UFSAR. This leads to a change in the special treatment requirements defined in the UFSAR, which had called for monthly surveillance testing. In applying the Integrated Decision Process in NEI 18-04 it was determined that the surveillance testing interval could be increased from monthly to quarterly while maintaining confidence that the reliability targets and design functions could be met. In applying Criterion (b) it is determined that there is no change to the risk significance of any LBE in which the pump provides a prevention or mitigation function. Therefore, Criterion (b) does not require prior NRC approval for the change to the special treatment.

4.3.3. Criterion (c)

(c) Result in more than a minimal increase in the consequences of a Design Basis Accident or any other event with which the plant is designed to cope described in the final safety analysis report (as updated). Commented [A61]: Same comment as above - this seems too broad.

Criterion (c) impacts NEI 18-04 DBAs only. Such DBAs are analyzed in a conservative manner Commented [A62]: This statement is inconsistent with the definition above, which includes more than DBAs.

against the same consequence criteria as in the current regulations, 10 CFR 50.34 and 10 CFR 100. Note that Criterion (c) addresses only consequences because the selection of DBAs in the LMP framework is not based on frequency of occurrence or risk but rather by application of deterministic rules by converting DBEs to DBAs while crediting only SR SSCs in the performance of Required Safety Functions. In NEI 18-04 and NEI 21-07 the consequences of DBAs are assessed in terms of public dose at the site boundary. The only activities affecting on-site dose consequences that may require prior NRC approval that are relevant to the LMP-based safety case are limited to operator actions that may be required to perform one or more Required Safety Functions as identified in Section 5.4 of the UFSAR per NEI 21-07. The onsite doses to consider in this respect are limited to those dose exposures during the time frames that the operator actions need to be performed to execute the Required Safety Functions. For changes affecting the dose to operators performing required actions outside the control room, an increase is considered more than minimal if the resultant mission dose exceeds applicable limits for 43

dose to control room operators The guidance in the remainder of this section applies to evaluation of effects of changes on main control room and off-site doses. Commented [A63]: As this guidance being technology inclusive, consideration of facilities with no control rooms (e.g.,

microreactors) may be needed in this area.

The consequences covered include dose resulting from any DBA evaluated in the UFSAR. The DBAs include those covered in UFSAR Chapter 3 and other events described in the UFSAR which the plant is designed to cope. Such events should be addressed by the NRCs ARCAP guidance. Note that such events do not include AOOs, DBEs, and BDBEs, which are addressed by Criterion (a). Commented [A64]: This sentence does not align with the criterion definition as currently written.

10 CFR Part 20 establishes requirements for protection against radiation during normal operations, including dose criteria relative to radioactive waste handling and effluents. DBA dose criteria and evaluation guidance provided in this document are not applicable to proposed activities governed by 10 CFR Part 20 requirements.

The dose consequences referred to in this guidance are those calculated by licensees, not the results of independent, confirmatory dose analyses by NRC that may be documented in SERs.

The evaluation should determine the dose that would likely result from DBAs associated with the proposed activity. If a proposed activity would result in more than a minimal increase in dose from the existing calculated dose for any DBA, then the activity would require prior NRC approval. Where a change in consequences is so small or the uncertainties in determining whether a change in consequences has occurred are such that it cannot be reasonably concluded that the consequences have actually changed (i.e., there is no clear trend toward increasing the consequences), the change need not be considered an increase in consequences.

Limits are established for radiation protection to permit access and occupancy of the control room under the most limiting conditions without personnel receiving radiation exposure in excess of 5 total effective dose equivalent (TEDE) for the duration of the event. These limits are Commented [A65]: Given in principal design criteria for non-LWRs. (e.g., See ARDC 19).

applicable only if control room access is required for mitigation of the event. 10 CFR Part 100 establishes requirements for exclusion area and low population zones around the reactor so that an individual located at any point on its the outer boundary of the zone would meet the relevant criteria set forth in 10 CFR 50.34(a)(1) for the radiological dose consequences of postulated accidents immediatelyfollowing onset of the postulatedfission product release would not receiveaTEDEinexcessof5rem. Commented [A66]: Power reactor siting, offsite accident dose criteria, and definition of EAB and LPZ are given in 10 CFR 100.21. It points to the criteria in 10 CFR 50.34 (2 rem TEDE for any 2-hr Therefore, for a given DBA, calculated or bounding dose values for that DBA would be period at the EAB, and 25 rem TEDE for the duration of the passage identified in the UFSAR. These dose values should be within the applicable regulatory limits. of the plume at any point on the outer boundary of the LPZ).

An increase in consequences from a proposed activity is defined to be no more than minimal if the increase (1) is less than or equal to 10% of the difference between the current calculated dose value and the regulatory limit. The current calculated dose values are those documented in the most up-to-date analyses of record.

In determining if there is more than a minimal increase in consequences, the first step is to determine which DBAs evaluated in the UFSAR may have their radiological consequences affected as a direct result of the proposed activity. Examples of questions that assist in this determination are:

Will the proposed activity change, prevent, or degrade the effectiveness of actions described or assumed in a DBA discussed in the UFSAR?

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 Will the proposed activity alter assumptions previously made in evaluating the radiological Commented [A67]: How does this interact with the screening and evaluation for a departure from a method of evaluation consequences of a DBA described in the UFSAR? (criterion (i))?

Will the proposed activity play a direct role in mitigating the radiological consequences of a DBA described in the UFSAR? Commented [A68]: Suggest bulletizing The next step is to determine if the proposed activity does, in fact, increase the radiological consequences of any of the DBAs evaluated in the UFSAR. If it is determined that the proposed activity does have an effect on the radiological consequences of any DBA analysis described in the UFSAR, then either:

Demonstrate and document that the radiological consequences of the DBA described in the UFSAR are bounding for the proposed activity (e.g., by showing that the results of the UFSAR analysis bound those that would be associated with the proposed activity), or Revise and document the analysis taking into account the proposed activity and determine if more than a minimal increase has occurred as described above.

The following examples illustrate the implementation of this criterion. In each example it is assumed that the calculated consequences do not include a change in the methodology for calculating the consequences. Changes in methodology would need to be separately considered under Criterion (i) as discussed in Section 4.3.9.

Example 1 The calculated dose to the control room operators following a loss of coolant DBA is 4 rem TEDE. A change is proposed to the control room ventilation system such that the calculated dose would increase to 4.5 rem. The regulations dictate that the control room doses are to be controlled to less than 5 rem. Although the new calculated dose is less than the regulatory limits, Commented [A69]: Related to Tom Kings comment above on the actual regulatory requirements in 10 CFR 50.34?

the incremental increase in dose (0.5 rem) exceeds the value of 10% of the difference between the previously calculated value and the regulatory value or 0.1 rem [10% of (5 rem - 4 rem)].

This change would require prior NRC review because the increase in consequences exceeds the minimal standard.

Example 2 The calculated public dose consequence for a particular steam generator tube rupture accident is 2 rem TEDE at the exclusion area boundary. As a result of a proposed change, the calculated dose consequence would increase to 3 rem. The increase is not more than minimal because the new calculated dose does not exceed the applicable guideline of 25 rem TEDE, nor does the incremental change in consequences (1 rem) exceed 10% of the difference between the previous calculated value and the regulatory limit of 25 rem TEDE. Ten percent of the difference between the regulatory limit (25 rem) and the calculated value (2 rem) is 2.3 rem (10% of 23).

Since 1 rem is less than 2.3 rem, this change does not cause more than a minimal increase in consequences therefore Criterion (c) does not require prior NRC approval for the change.

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4.3.4. Criterion (d)

(d) Result in identifying one or more AOO, DBE, or BDBE that is (i) not previously evaluated in the UFSAR and (ii) classified as risk significant according to NEI 18-04 LBE risk significance criteria.

Potential changes that could introduce newly identified LBEs are addressed by LMP 50.59 Commented [A70]: LMP does not have 50.59 criteria? Please clarify.

Criterion (d). Such changes should be evaluated by revisiting the NEI 18-04 process for identifying LBEs.

The evaluation of the change should determine whether the change alters the event sequence plant response model in a manner that introduces a new event sequence or event sequence family. It is important to note that Criterion (d) is satisfied only when any new LBEs exceed the risk significance criteria in NEI 18-04 based on mean values of frequency and consequence. If a newly identified LBE is not risk significant, it has no material impact on the safety analysis basedonthemethodologyendorsedinRG1.233LMP-basedaffirmativesafetycase,sothechangewouldnotrequirepriorNRC review. Also, because a new DBA would require a new DBE, Criterion (d) implicitly covers DBAs as well as AOOs, DBEs, and BDBEs.

The following examples illustrate the implementation of this criterion.

Example 1 The station desires to install a new process heat plant to make commercial quantities of hydrogen. The proposed location of the process heat plant for making the hydrogen is close enough for blast effects to interrupt normal operation and potentially damage equipment relied upon for normal cooling. The modification is in the conceptual stage; however, the change would likely result in one or more additional risk significant LBEs that are not evaluated in the UFSAR and therefore would require prior NRC approval per Criterion (d).

Example 2 A pump that is classified as NSRST is equipped with an automatic start feature. The automatic start feature is credited in the analyses of several LBEs, none of which are DBAs. A change is being considered to incorporate a new action to manually start the pump in lieu of the automatic Commented [A71]: The addition of safety significant operator actions should also enter this into the change process.

start feature. The proposed activity will revise operating procedures to incorporate the new action and implement a design change to delete the circuitry of the automatic start feature. The PRA model was revised for the purposes of analyzing the proposed activity. The change resulted in minimal impacts to the risk profile and all the change evaluation criteria were met. In this instance, the results indicate multiple events are changed, new LBEs are identified, and one existing LBE is changed to risk-significant. The Integrated Decision-Making Process (IDP) reviewed against the cumulative risk criterion and whether a new DBA should be declared and also determined that the impacted NSRST SSCs need not be SR but would benefit from different special treatment. Therefore, the proposed activity would require prior NRC approval per Criterion (d).

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 4.3.5. Criterion (e)

(e) Result in a change to any of the NEI 18-04 cumulative risk metrics that exceeds the cumulative risk targets in Section 3.3.5 of NEI 18-04.

Criterion (e) addresses the impact of a proposed change to cumulative risk. The cumulative risk metrics are defined in NEI 18-04 Section 3.3.5 and would be documented in the updated FSAR in accordance with NEI 21-07 SAR Section 4.1. This criterion provides confidence that an accumulation of changes over time, each of which is acceptable from an individual LBE perspective, does not lead to unacceptable cumulative risk.

The following example illustrates the implementation of this criterion.

Example 1 A component used to circulate primary coolant during normal operation is degrading. Plant procedures limit the components out of service time to no more than seven days before requiring a unit shutdown. However, the maintenance is expected to take longer than seven days and the plant is evaluating the option to stay online to support increased demands on the power grid. The proposed activity is to extend the procedurally allowed out of service time before requiring a plant shutdown. An analysis was conducted by extending the components out of service time in the PRA model. The purpose of the analysis was to evaluate the impact of extending the allowed out of service time to accomplish the repair during normal operation. The study concluded that the proposed activity would not result in a change that exceeds the cumulative risk targets in Section 3.3.5 of NEI 18-04. Therefore, Criterion (e) does not require prior NRC approval for the change. Commented [A72]: It should be written differently such as:

The proposed activity does not require 4.3.6. Criterion (f)

(f) Result in an increase in the frequency and/or consequences of a malfunction of any safety-significant SSC that would change the classification of the SSC from non-risk significant to risk-significant.

Criterion (f) is based on NEI 18-04 SSC safety classification. Changes that may impact SSC risk significance or safety classification should be evaluated by revisiting the pertinent processes in NEI 18-04 Section 3.2.2 after quantifying the impact on risk using the PRA.

The following example illustrates the implementation of this criterion.

Example 1 A pump that is classified as NSRST is equipped with an automatic start feature. The automatic start feature is credited in the analyses of several LBEs, none of which are DBAs. A change is being considered to incorporate a new action to manually start the pump in lieu of the automatic Commented [A73]: Addition of safety-significant operator actions should immediately enter the change into the screening start feature. The proposed activity will revise operating procedures to incorporate the new process, as well as require an HFE justification.

action and will implement a design change to delete the circuitry of the automatic start feature.

The PRA model was revised for the purposes of analyzing the proposed activity; the change did 47

not result in the identification of any new risk significant SSCs. Therefore, Criterion (f) does not require prior NRC approval for the change. Commented [A74]: It should be written differently such as:

The proposed activity does not require 4.3.7. Criterion (g)

(g) Result in a change in Safety-Related SSC classification.

It is possible that a change might lead to a reclassification of an SSC from NST or NSRST to SR, although such changes are expected to be rare. For example, if a change introduces a new hazard that was not addressed in the existing classification, it may be determined that it is more cost effective to select and reclassify an existing NST or NSRST SSC to SR to perform a Required Safety Function to protect against that hazard than to impose new design criteria on an existing SR SSC to protect against that hazard. Commented [A75]: What about changing the classification from SR to NSTST or NST? This example should also fit within this criterion.

In other situations, for certain changes, it may be necessary to reclassify an SSC from NST to NSRST to maintain the adequacy of DID. If the affected SSC was originally NST and the SSC is subsequently reclassified as NSRST then Criterion (g) does not require prior NRC approval for the change.

The following example illustrates the implementation of this criterion.

Example 1 An air-cooling system is classified as NST. A proposed activity is being considered to incorporate a debris filtration screen on the systems intake to reduce the potential for internal fouling. However, the screen would result in an increased potential for flow restriction during some postulated events. The proposed modification is modeled in the PRA, and it is determined that the new failure mode (i.e., clogged screen) would cause the system to less reliable and would result in a change to the systems safety classification from NST to NSRST. Therefore, Criterion (g) does not require prior NRC approval for the change. Commented [A76]: It should be written differently such as:

The proposed activity does not require 4.3.8. Criterion (h)

(h) Result in a change to the performance of a safety-significant SSC that would adversely change the defense-in-depth adequacy determination.

Criterion (h) addresses adverse effects on DID adequacy that change the DID adequacy determination. DID has an important, formalized role in the LMP-based affirmative safety case as addressed in Chapter 5 of NEI 18-04 and Chapter 4 of NEI 21-07. Elements of the DID baseline may relate to plant capability (e.g., SSCs) or be programmatic in nature (e.g., testing).

DID elements will vary among technology types, specific designs, and the nature of the safety case, so the DID baseline for one reactor may be very different from another.

NEI 18-04 Section 5.9.6 sets forth specific questions associated with the consideration of plant changes, repeated below.

  • Does the change introduce a new LBE for the plant?

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07

  • Does the change increase the risk of LBEs previously considered to be of no/low risk significance to the point that it will be considered risk-significant after the change is made?
  • Does the change reduce the number of layers of defense for any impacted LBEs or materially alter the effectiveness of an existing layer of defense?
  • Does the change significantly increase the dependency on a single feature relied on in risk-significant LBEs?

The first and second bullets are addressed by Criteria (a) and (b), respectively. The third and fourth bullets relate directly to DID and are addressed by Criterion (h).

NEI 18-04 Section 5.9.3 describes how DID adequacy is confirmed in establishing the DID baseline. NEI 21-07 specifies that SAR Section 4.2.3 will document the integrated DID evaluation by addressing the confirmatory DID criteria in NEI 18-04 Section 5.9.3. Thus, the focus of the evaluation of the effect of a proposed change on DID adequacy per Criterion (h) should be on the integrated DID evaluation as documented in SAR Section 4.2.3, which addresses both plant capability and programmatic DID.

Any changes which would change the defense-in-depthDID adequacy determination would require prior NRC approval. In this instance, the adverse effect of the change relates to the design function the SSC is intended to accomplish in support of DID adequacy. Some of the confirmatory DID criteria are amenable to quantitative assessment (e.g., performance targets for SSC reliability and capability are identified), while others require a qualitative evaluation (e.g.,

prevention/mitigation balance is sufficient). Given the potential for variability in the DID baselines for different designs, it is not practical for guidance to specify, in advance, finite change control acceptance criteria for all considerations related to DID adequacy. The evaluation will be based on the DID baseline information in NEI 21-07 SAR Section 4.2.3 and the design records.

The nature of the change and its impact on the LMP-based affirmative safety case will impact the approach taken to carrying out the DID portion of the change evaluation. It is anticipated that many changes will be simple and limited in scope such that the evaluation against LMP 50.59 Criterion (h) will be relatively straightforward, using the information and criteria documented in Commented [A77]: LMP does not have 50.59 criterion? Please clarify.

the SAR and the plant records. However, some changes may require a more comprehensive Integrated Decision-Making Process IDP review of DID adequacy, including the possibility of utilizing an Integrated Decision-Making Process Panel (IDPP), as described in NEI 18-04 Chapter 5.

Examples of situations in which an integrated evaluation of DID using an IDP may be appropriate are:

1. The change reduces the number of layers of defense for any impacted LBEs or the change materially alters the effectiveness of an existing layer of defense.
2. The change materially alters the prevention-mitigation balance across all layers of defense such that an over-dependance on a single layer results.

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3. The change significantly increases the dependency on a single feature relied on in risk-significant LBEs.
4. The sources of uncertainty from changes in SSC performance increase to the point that the impact of the change across all affected LBEs dominate the uncertainty evaluation. This is described in NEI 18-04 Section 4.2.2.

Consistent with Section 4.2, the licensee may include additional compensatory elements in the Commented [A78]: Of NEI 18-04?

change in order to maintain DID adequacy and preclude the need for prior NRC approval. The use of additional compensatory actions to reduce the risk-impacts should be reviewed by an IntegratedDecision-MakingProcessPanel(IDPP). DIDisinterwovenwithmanypartsoftheLMP-basedaffirmativesafetycase,anditis important to provide licensees with flexibility to address it holistically provided overall DID adequacy is maintained.

Once the Criterion (h) determination is made, the basis for the determination must be documented as discussed in Section 5. In addition to updating plant records, if necessary, there should be an update of the DID baseline evaluation in the UFSAR.

4.3.8.1. Examples of evaluation of DID adequacy The following examples illustrate the implementation of this criterion.

Example 1 A change is being proposed that increases the reliability of a function by installing a cross-connection that improves normal operation reliability and flexibility but reduces independence between multiple trains of a system for a single LBE that is not risk significant. The design function of the system is potentially adversely eaffected since change may result in altered capability of the function depending on any new failure modes or reduced flows for the introduced from the cross-connected system operation. The additional layers of defense that are available for this event should be evaluated for their sufficiency. Depending on the IDP evaluation of whether this change created the need for additional NSRST SSC requirements, some additional special treatment may be needed. Since the original FSAR included this LBE, the changes in the baseline described in the UFSAR would need to be updated. Assuming the results of the analysis show that DID adequacy is maintained following the change, the proposed change does not require NRC pre-approval per Criterion (h). Commented [A79]: Consider adding "Conversely if the analysis shows that additional special treatments may be needed to maintain DID adequacy, then NRC pre-approval would be required The overall plant performance change is shown in Error! Reference source not found.Figure 4. per this criterion."

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 Figure 4. Overall Plant Performance Change in Example 1 Example 2 In many events, non-safety related reactor system heat rejection is a single feature that is present across multiple events and across layers of defense for a given event sequence family.

Programmatic special treatment may be provided for portions of the manufacturing, construction and operational phases and should be noted in the baseline DID adequacy evaluation in the UFSAR. A reduction in test frequency or scope should require an IDP evaluation to consider the integrated impacts on cumulative uncertainties, such as increased failure probabilities, the nature of the failure mechanism and the adequacy of the remaining special treatments to minimize the likelihood of functional failures. If the change invalidates any of the DID adequacy attributes, the change should be rejected by the IDP and reconsidered. If DID adequacy is maintained, Criterion (h) does not require prior NRC approval for the change. Commented [A80]: It should be written differently such as:

The proposed activity does not require Example 3 A valve is classified as NSRST because it prevents reliance on a single element of design to perform a Required Safety Function in applying the plant capability DID criteria in Table 5-2 in NEI 18-04. The valve does not meet the SSC risk significance criteria in NEI 18-04. During surveillance tests a new degradation mechanism is identified that is determined to be mitigated by revised maintenance procedures for inspection and cleaning of the component. This results in developing a change to the special treatment requirements that compensates for any adverse 51

impacts on the reliability of the valve. Therefore, DID adequacy is maintained by implementing the change. Criterion (h) does not require prior NRC approval for the change. Commented [A81]: It should be written differently such as:

The proposed activity does not require 4.3.9. Criterion (i)

(i) Result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses.

The UFSAR contains design and licensing basis information for a nuclear power facility, including description on how regulatory requirements for design bases are met and how the facility responds to various DBAs.

The scope of the methods considered in this criterion include the methods used to evaluate the consequences of the DBAs and to confirm that the Safety-Related Design Criteria for SR SSCs are met. This includes the methods that are used to determine that the SR SSCs are capable of performing their Required Safety Functions for all the DBAs including those resulting from an internal or external hazard at their respective DBHLs. Commented [A82]: This scope seems too limiting, as discussed in multiple previous comments. Design basis functions cover SSCs that perform Required Safety Functions, risk significant functions, Analytical methods are a fundamental part of demonstrating how the design meets regulatory or functions needed for adequate Defense-in-Depth.

requirements and that the facility response to accidents and events is acceptable. As such, in cases where the analytical methodology was considered to be an important part of the conclusion that the facility met the required design bases, these analytical methods were described in the UFSAR and received varying levels of NRC review and approval during licensing.

Because this guidance provides a process for determining if prior NRC approval is required before making changes to the facility as described in the UFSAR, changes to the methodologies described in the UFSAR also fall under the provisions of this process, specifically Criterion (i).

In general, licensees can make changes to elements of a methodology without first obtaining a license amendment if the results are essentially the same as, or more conservative than, previous results. Similarly, licensees can also use different methods without first obtaining a license amendment if those methods have been approved by NRC for the intended application.

If the proposed activity does not involve a change to a method of evaluation, then the activity evaluation should reflect that this criterion is not applicable. If the activity involves only a change to a method of evaluation, then the activity evaluation should reflect that Criteria (a) - (h) are not applicable.

The first step in applying this criterion is to identify the methods of evaluation that are affected by the change. This is accomplished during application of the screening criteria in Section 4.2.1.3.

Next, the licensee must determine whether the change constitutes a departure from a method of evaluation that would require prior NRC approval. As discussed further below, for purposes of evaluations under this criterion, the following changes are considered a departure from a method of evaluation described in the UFSAR:

  • Changes to any element of analysis methodology that yield results that are nonconservative or not essentially the same as the results from the analyses of record

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07

  • Use of new or different methods of evaluation that are not approved by NRC for the intended application By way of contrast, the following changes are not considered departures from a method of evaluation described in the UFSAR:
  • Departures from methods of evaluation that are not described, outlined, or summarized in the UFSAR or are controlled by an alternative regulation. (Such changes may have been screened out as discussed in Sections 4.1.1 and 4.2.1.3).
  • Use of a new NRC-approved methodology (e.g., new or upgraded computer code) to reduce uncertainty, provide more precise results or other reason, provided such use is (a) based on sound engineering practice, (b) appropriate for the intended application and (c) within the limitations of the applicable SER. The basis for this determination should be documented in the licensee evaluation.
  • Use of a methodology revision that is documented as providing results that are essentially the same as, or more conservative than, either the previous revision of the same methodology or another methodology previously accepted by NRC through issuance of an SER.

Note that changes in evaluation methods used in the PRA, including those used for AOOs, DBEs, and BDBEs, do not normally require prior NRC approval because they are addressed through adherence to ASME/ANS RA-S-1.4-2021 (see Section 4.1.7). However, if a method of evaluation used in the PRA is also described in the UFSAR for application to an analysis that demonstrates that DBA limits are met for SSCs that support the retention of radioactive material than the method of evaluation would require prior NRC approval if Criterion (i) is met.

Section 4.3.9.1 provides guidance for making changes to one or more elements of an existing method of evaluation used to establish the design bases or in the safety analyses. Section 4.3.9.2 provides guidance for adopting an entirely new method of evaluation to replace an existing one.

Examples illustrating the implementation of this criterion are provided in Section 4.3.9.3.

4.3.9.1.1. Guidance for Changing One or More Elements of a Method of Evaluation Commented [A83]: Should be labeled 4.3.9.1 (vs 4.3.9.1.1)?

The definition of departure provides licensees with the flexibility to make changes to methods of evaluation whose results are conservative or that are not important to the demonstrations of performance that the analyses provide. Changes to elements of analysis methods that yield conservative results or results that are essentially the same would not be departures from approved methods.

Conservative vs. Nonconservative Results Gaining margin by changing one or more elements of a method of evaluation is considered to be a nonconservative change and thus a departure from a method of evaluation. Such departures require prior NRC approval of the revised method. Analytical results obtained by changing any element of a method are conservative relative to the previous results, if they are closer to design bases limits or safety analyses limits (e.g., applicable acceptance guidelines). For example, a change from 45 psig to 48 psig in the result of a containment peak pressure analysis 53

(with design basis limit of 50 psig) using a revised method of evaluation would be considered a conservative change when applying this criterion. In other words, the revised method is more conservative if it predicts more severe conditions given the same set of inputs. This is because results closer to limiting values are considered conservative in the sense that the new analysis result provides less margin to applicable limits for making potential physical or procedure changes without a license amendment.

In contrast, if the use of a modified method of evaluation resulted in a change in calculated containment peak pressure from 45 psig to 40 psig, this would be a nonconservative change.

That is because the change would result in more margin being available (to the design basis limit of 50 psig) for the licensee to make more significant changes to the physical facility or procedures.

Essentially the Same Licensees may change one or more elements of a method of evaluation such that results move in the nonconservative direction without prior NRC approval, provided the revised result is essentially the same as the previous result. Results are essentially the same if they are within the margin of error for the type of analysis being performed. Variation in results due to routine analysis sensitivities or calculational differences (e.g., rounding errors and use of different computational platforms) would typically be within the analysis margin of error and thus considered essentially the same. For example, when a method is applied using a different computational platform (mainframe vs. workstation), results of cases run on the two platforms differed by less than 1%, which is the margin of error for this type of calculation. Thus, the results are essentially the same, and do not constitute a departure from a method that requires prior NRC approval.

The determination of whether a new analysis result would be considered essentially the same as the previous result can be made through benchmarking the revised method to the existing one or may be apparent from the nature of the differences between the methods. When benchmarking a revised method to determine how it compares to the previous one, the analyses performed must be for the same set of plant conditions to ensure that the results are comparable. Comparison of analysis methods should consider both the peak values and time behavior of results, and engineering judgment should be applied in determining whether two methods yield results that are essentially the same.

4.3.9.2. Guidance for Changing from One Method of Evaluation to Another The definition of departure provides licensees with the flexibility to make changes by implementing this guidance from one method of evaluation to another provided that the new method is approved by NRC for the intended application. A new method is approved by NRC for intended application if it is approved for the type of analysis being conducted, and applicable terms, conditions and limitations for its use are satisfied.

NRC approval has typically followed one of two paths. Most reactor or fuel vendors and several utilities have prepared and obtained NRC approval of topical reports that describe methodologies for the performance of a given type or class of analysis. Through an SER, NRC approved the use of the methodologies for a given class of power plants. In some cases, NRC has accorded

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 generic approval of analysis methodologies. Terms, conditions, and limitations relating to the application of the methodologies are usually documented in the topical reports, the SER, and correspondence between NRC and the methodology owner that is referenced in the SER or associated transmittal letter.

The second path is the approval of a specific analysis rather than a more generic methodology.

In these cases, NRCs approval has typically been part of a plants licensing basis and limited to a given plant design and a given application. Again, a thorough understanding of the terms, conditions and limitations relating to the application of the methodology is essential. This information is usually documented in the original license application or license amendment request, the SER, and any correspondence between NRC and the analysis owner that is referenced in the SER or associated transmittal letter.

It is incumbent upon the users of a new methodologyeven one generically approved by NRCto ensure they have a thorough understanding of the methodology in question, the terms of its existing application and conditions/limitations on its use. A range of considerations is identified below that may be applicable to determining whether new methods are technically appropriate for the intended application. The licensee should address these and similar considerations, as applicable, and document in the activity evaluation the basis for determining that a method is appropriate and approved for the intended application. To obtain an adequate understanding of the method and basis for determining it is approved for use in the intended application, licensees should consult various sources, as appropriate. These include SERs, topical reports, and licensee correspondence with NRC and licensee personnel familiar with the existing application of the method. If adequate information cannot be found on which to base the intended application of the methodology, the method should not be considered approved by NRC for the intended application.

The applicable terms and conditions for the use of a methodology are not limited to a specific analysis; the qualification of the organization applying the methodology is also a consideration.

Through Generic Letter 83-11, Supplement 1,9 NRC has established a method by which licensees can demonstrate they are generally qualified to perform safety analyses. Licensees thus qualified can apply methods that have been reviewed and approved by NRC or that have been otherwise accepted as part of another plants licensing basis, without requiring prior NRC approval. Licensees that have not satisfied the guidelines of Generic Letter 83-11, Supplement 1, may, of course, continue to seek plant-specific approval to use new methods of evaluation.

When considering the application of a methodology, it is necessary to adopt the methodology in toto and apply it consistent with applicable terms, conditions, and limitations. Mixing attributes of new and existing methodologies is considered a revision to a methodology and must be evaluated as such per the guidance in Section 4.3.9.1.

9 Generic Letter 83-11, Supplement 1, Licensee Qualification for Performing Safety Analyses, June 24, 1999.

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Considerations for Determining if New Methods May be Considered Approved by NRC for the Intended Application The following questions highlight important considerations for determining that a particular application of a different method is technically appropriate for the intended application, within the bounds of what has been found acceptable by NRC and does not require prior NRC approval.

  • Is the application of the methodology consistent with the facilitys licensing basis (e.g.,

plant-specific commitments)? Will the methodology supersede a methodology addressed by other regulations such as 10 CFR 50.46, 10 CFR 50.55a or the plant technical specifications Commented [A84]: 50.46 is not applicable to non-LWRs.

(Core Operating Limits Report or Pressure/Temperature Limits Report)? Is the methodology consistent with relevant industry standards?

  • Does the application of the new methodology require an exemption(s) from regulations or plant-specific commitments, exceptions to relevant industry standards and guidelines, or is otherwise inconsistent with a facilitys licensing basis? If so, then prior NRC approval may be required. The applicable change process must be followed to make the plants licensing basis consistent with the requirements of the new methodology.
  • Does the new method involve the use of a computer code, and if so, has the code been installed in accordance with applicable software quality assurance requirements? Has the plant-specific model been adequately qualified through benchmark comparisons against test data, plant data or approved engineering analyses? Is the application consistent with the capabilities and limitations of the computer code? Has industry experience with the computer code been appropriately considered?
  • Is the computer code installation and plant-specific model qualification directly transferable from one organization to another? Has the installation and qualification been performed in accordance with the quality assurance program?
  • Is the facility for which the methodology has been approved designed and operated in the same manner as the facility to which the methodology is to be applied? Is the relevant equipment the same? Does the equipment have the same pedigree (e.g., Class 1E, Seismic Category I, etc.)? Are the relevant failure modes and effects analyses the same? If the plant is designed and operated in a similar, but not identical, manner, the following types of considerations should be addressed to assess the applicability of the methodology:

How could those differences affect the methodology?

Are additional sensitivity studies required?

Should additional single failure scenarios be considered?

Are analyses of limiting scenarios, effects of equipment failures, etc., applicable for the specific plant design?

Can analyses be made while maintaining compliance with both the intent and literal definition of the methodology? Commented [A85]: Suggest sub-bullitizing Differences in the plant configurations and licensing bases could invalidate the application of a particular methodology. For example, the licensing basis of older vintage plants may not include an analysis of the feedwater line break event that is required in later vintage plants. Some plants may be required to postulate a loss of off-site power or a maximum break size for certain events;

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07 others may have obtained exemptions to these requirements from NRC. Some plants may have pressurizer power-operated relief valves that are qualified for water relief; other plants do not.

Plant specific failure modes and effects analyses may reveal new potential single failure scenarios that cannot be adequately assessed with the original methodology. The existence of these differences does not preclude application of a new methodology to a facility; however, differences must be identified, understood and the basis documented for concluding that the differences are not relevant to determining that the new application is technically appropriate.

4.3.9.3. Examples of evaluation of changes to or departures from methods The following examples illustrate the implementation of this criterion.

Example 1 The UFSAR states that a damping value of 0.5% is used in the seismic analysis of safety-related piping. The licensee wishes to change this value to 2% to reanalyze the seismic loads for the piping. Using a higher damping value to represent the response of the piping to the acceleration from the postulated earthquake in the analysis would result in lower calculated stresses because the increased damping reduces the loads. Since this analysis was used in establishing the seismic design bases for the piping, and since this is a change to an element of the method that is not conservative and is not essentially the same, this change would require prior NRC approval under this criterion. Commented [A86]: This example is confusing. Earlier in this document it is stated that changes to inputs would NOT be a change in method of evaluation. Here, an input value is being On the other hand, had NRC approved an alternate method of seismic analysis that allowed 2% changed but not the actual method and yet it states that this damping provided certain other assumptions were made, and the licensee used the complete set change would require prior NRC approval.

of assumptions to perform its analysis, then the 2% damping under these circumstances would not be a departure because this method of evaluation is considered approved by NRC for the intended application.

Example 2 A facility has a design basis containment pressure limit of 50 psig. The current worst-case design basis accident calculation results in a peak pressure of 45 psig within two minutes. The licensee revises the method of evaluation, and the recalculated result is 40 psig. This change would require prior NRC approval because the result of the recalculation is not conservative. If the licensee used a different method that was approved by NRC and met all the terms and conditions of the method, a recalculated result of 40 psig would not require prior NRC approval.

Example 3 A licensee revises the seismic analysis described in the UFSAR to include an inelastic analysis procedure. This revised method is used to demonstrate that cable trays have greater capacity than previously calculated. This change would require prior NRC approval as it would not produce results that are essentially the same.

Example 4 Licensee X has received NRC approval for the use of a method of evaluation at Facility A for performing steamline break mass and energy release calculations for environmental qualification evaluations. The terms and conditions for the use of the method are detailed in the NRC SER.

57

The SER also describes limitations associated with the method. Licensee Y wants to apply the method at its Facility B. Licensee Y has satisfied the guidelines of GL 83-11, Supplement 1.

After reviewing the method, approved application, SER, and related documentation, to verify that applicable terms, conditions, and limitations are met and to ensure the method is applicable to their type of plant, Licensee Y conducts an activity evaluation. Licensee Y concludes that the change is not a departure from a method of evaluation because it has determined the method is appropriate for the intended application, the terms and conditions for its use as specified in the SER have been satisfied, and the method has been approved by NRC.

Example 5 NRC has approved the use of computer code and the associated analysis of a steamline break for use in the evaluation of component stresses. A licensee uses the same computer code and analysis methodology to replace its evaluation of the containment temperature response. This change would require prior NRC approval unless the methodology had been previously approved for evaluating containment temperature response.

Example 6 A new version of the computer software is available that is used for modeling the peak fuel and cladding temperature as part of the DBA safety analyses. The new version of the software more accurately models the physical phenomena inside the reactor core and results in peak temperatures being lower than currently analyzed. The proposed method of analysis is considered nonconservative as it gains margin to the safety limits for the fuel and cladding.

Therefore, prior NRC approval is required to adopt the new software before the results can be incorporated into the analyses of record.

Example 7 A new version of the computer software is available that is used for modeling the peak fuel and cladding temperature as part of the analyses for LBEs that are not a DBAs. The new version of the software more accurately models the physical phenomena inside the reactor core and results in peak temperatures being lower than currently analyzed. The proposed activity is considered a change to a PRA-related method of evaluation (i.e., it is not associated with DBA analyses).

Therefore, the activity will not require prior NRC approval as part of this change control process.

4.4. Utilizing an Activity Evaluation to Assess Compensatory Actions to Address Nonconforming or Degraded Conditions Three general courses of action are available to licensees to address non-conforming and degraded conditions. Whether or not this guidance must be applied, and the focus of an activity evaluation if one is required, depends on the corrective action plan chosen by the licensee, as discussed below.

  • If the licensee intends to restore the SSC back to its as-designed condition then this corrective action should be performed in accordance with 10 CFR Part 50, Appendix B (i.e.,

in a timely manner commensurate with safety). This activity is not subject to an activity evaluation.

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07

  • If an interim compensatory action is taken to address the condition and involves a temporary procedure or facility change, this guidance should be applied to assess the temporary change.

The intent is to determine whether the temporary change/compensatory action itself (not the degraded condition) impacts other aspects of the facility or procedures described in the UFSAR. In considering whether a temporary change impacts other aspects of the facility, a licensee should pay particular attention to ancillary aspects of the temporary change that result from actions taken to directly compensate for the degraded condition.

  • If the licensee corrective action is either to accept the condition as-is resulting in something different than its as-designed condition, or to change the facility or procedures, an activity evaluation should be performed for the proposed corrective action, unless another regulation applies, e.g., 10 CFR 50.55a. In these cases, the final corrective action becomes the proposed change that would be subject to an activity evaluation.

In resolving degraded or nonconforming conditions, the need to obtain NRC approval for a proposed activity does not affect the licensee's authority to operate the plant. The licensee may make mode changes, restart from outages, etc., provided that necessary SSCs are operable and the degraded condition is not in conflict with the technical specifications or the license.

The following example illustrates the process for implementing a temporary change as a compensatory action to address a degraded/nonconforming condition.

Example A level transmitter for one Reactor Coolant Pump (RCP) lower oil reservoir failed while at power. The transmitter provides an alarm function, but not an automatic protective action function. The transmitter and associated alarm are described in the UFSAR, as protective features for the RCPs, but no technical specification applies. Loss of the transmitter does not result in the loss of operability for any technical specification equipment. The transmitter fails in a direction resulting in a continuous alarm in the control room. The alarm circuitry provides a common alarm for both the upper and lower oil reservoir circuits, so transmitter failure causes a hanging alarm and a masking of proper operation of the remaining functional transmitter.

Precautionary measures are taken to monitor lower reservoir oil level as outlined in the alarm manual using available alternate means. An interim compensatory action is proposed to lift the leads (temporary change) from the failed transmitter to restore the alarm function for the remaining functioning transmitter.

Lifting the leads is a compensatory action (temporary change) that is subject to an activity evaluation. The activity screening would be applied to the temporary change itself (lifted leads),

not the degraded condition (failed transmitter), to determine its impact on other aspects of the facility described in the UFSAR. If screening determines that no other UFSAR-described SSCs would be affected by this compensatory action, the temporary change would screen out, i.e., not require an activity evaluation.

4.5. Disposition of Activity Evaluations There are two possible conclusions to an activity evaluation:

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The proposed activity may be implemented without prior NRC approval.

The proposed activity requires prior NRC approval. Commented [A87]: Suggest bulletizing Where an activity requires prior NRC approval, the activity must be approved by NRC via license amendment in accordance with 10 CFR 50.90 prior to implementation. An activity is considered implemented when it provides its intended function, that is, when it is placed in service and declared operable. Thus, a licensee may design, plan, install and test a modification prior to receiving the license amendment to the extent that these preliminary activities do not themselves require prior NRC approval.

For example, a modification to a facility involved the replacement of a train of a safety system with one including diverse primary components (diesel-driven pump vice a motor-driven pump).

The installation of the replacement train was largely in a new, separate structure. Ultimately the modification would require NRC approval because of impacts on the technical specifications as well as due to differences in reliability of the replacement pump in some situations. There was insufficient time to seek and gain NRC approval prior to construction. The licensee prepared an activity screening to support construction of the separate structure through preliminary testing.

The limited interfaces with the existing facility were assessed and determined to not change the facility or procedures as described in the UFSAR. Upon receipt of the license amendment the final tie-in, testing and operation were fully authorized. This guidance should be applied to any aspects of the activity not adequately addressed in the license amendment request and/or associated SER.

For proposed activities that are determined to require prior NRC approval, there are three possible options:

Cancel the planned activity.

Redesign the proposed activity so that it may proceed without prior NRC approval.

Apply for and obtain a license amendment under 10 CFR 50.90 prior to implementing the activity. Technical and licensing evaluations performed for such activities may be used as part of the basis for license amendment requests. Commented [A88]: Suggest bulletizing It is important to remember that determining that a proposed activity requires prior NRC approval does not determine whether it is safe. In fact, a proposed activity that requires prior NRC approval may significantly enhance overall plant safety at the expense of a small adverse impact in a specific area. It is the responsibility of the utility to assure that proposed activities are safe, and it is the role of NRC to confirm the safety of those activities that are determined to require prior NRC review.

TIRICE for Non-Light Water Reactors Industry Guidance for the Evaluation of Changes to a Facility or Procedures, and Conducting Tests or Experiments Applicable to Facilities Utilizing NEI 18-04 and NEI 21-07

5. DOCUMENTATION AND REPORTING The following documentation and recordkeeping are required.

The licensee shall maintain records of changes in the facility, of changes in procedures, and of tests and experiments made pursuant to paragraph (c) of this section. These records must include Commented [A89]: There is no paragraph (c) a written evaluation that provides the bases for the determination that the change, test or experiment does not require a license amendment pursuant to paragraph (c)(2) of this section. Commented [A90]: There is no paragraph (c)

The licensee shall submit, as specified in 10 CFR Part 50.4, a report containing a brief description of any changes, tests, and experiments, including a summary of the evaluation of each. A report must be submitted at intervals not to exceed 24 months.

The records of changes in the facility must be maintained until the termination of a license issued pursuant to this part or the termination of a license issued pursuant to 10 CFR Part 54, whichever is later. Records of changes in procedures and records of tests and experiments must be maintained for a period of 5 years.

The documentation and reporting requirements contained in this guidance apply to activities that require evaluation against the nine criteria of Sections 4.3.1 through 4.3.9 and are determined not to require prior NRC approval. This pertains to those activities that were evaluated against the nine evaluation criteria (because, for example, they affect the facility as described in the UFSAR), but not to those activities or changes that were screened out. Similarly, documentation and reporting specified by this guidance is not required for activities that are canceled or that are determined to require prior NRC approval and are implemented via the license amendment request process.

Documenting Activity Evaluations In performing an activity evaluation of a proposed activity, the evaluator must address the nine criteria in Sections 4.3.1 through 4.3.9 to determine if prior NRC approval is required. Although the conclusion in each criterion may be simply yes, no, or not applicable, there must be an accompanying explanation providing adequate basis for the conclusion. These explanations should be complete in the sense that another knowledgeable reviewer could draw the same conclusion. Restatement of the criteria in a negative sense or making simple statements of conclusion is not sufficient and should be avoided. It is recognized, however, that for certain very simple activities, a statement of the conclusion with identification of references consulted to support the conclusion would be adequate and the activity evaluation could be very brief.

The importance of the documentation is emphasized by the fact that experience and engineering knowledge (other than models and experimental data) are often relied upon in determining whether evaluation criteria are met. Thus, the basis for the engineering judgment and the logic used in the determination should be documented to the extent practicable and to a degree commensurate with the safety significance and complexity of the activity. This type of documentation is of particular importance in areas where no established consensus methods are available, such as for software reliability, or the use of commercial-grade hardware and software where full documentation of the design process is not available.

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Since an important goal of the activity evaluation is completeness, the items considered by the evaluator must be clearly stated.

Each activity evaluation is unique. Although each applicable criterion must be addressed, the questions and considerations listed throughout this guidance to assist evaluating the criteria are not requirements for all evaluations. Some evaluations may require that none of these questions be addressed while others will require additional considerations beyond those identified in this guidance.

When preparing activity evaluations, licensees may combine responses to individual criteria or reference other portions of the evaluation.

As discussed in Section 4.2.3, licensees may elect to use screening criteria to limit the number of activities for which written activity evaluations are performed. A documentation basis should be maintained for determinations that the changes meet the screening criteria, i.e., screen out. This documentation does not constitute the record of changes required as defined in this guidance, and thus is not subject to the recordkeeping requirements of the rule.

Reporting to NRC A summary of activity evaluations for activities implemented under this guidance must be provided to NRC. Activities that were screened out, canceled, or implemented via license amendment need not be included in this report. The reporting requirement (every 24 months) is identical to that for UFSAR updates such that licensees may provide these reports to NRC on the same schedule. Commented [A91]: Suggest adding even page numbers missing in this doc.

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