TSTF-24-07, PRM-50-126 - Petition for Rulemaking to Amend 10 CFR 50.36, Technical Specifications from Brian D. Mann on Behalf of the Technical Specifications Task Force

PRM-50-126 - Petition for Rulemaking to Amend 10 CFR 50.36, Technical Specifications from Brian D. Mann on Behalf of the Technical Specifications Task Force

ML24274A209
Person / Time
Site:	Technical Specifications Task Force
Issue date:	09/13/2024
From:	Mann B Technical Specifications Task Force
To:	NRC/SECY/RAS
References
TSTF-24-07
Download: ML24274A209 (1)

Text

11921 Rockville Pike, Suite 100, Rockville, MD 20852 Phone: 804-339-7034 Administration by EXCEL Services Corporation TECHNICAL SPECIFICATIONS TASK FORCE A JOINT OWNERS GROUP ACTIVITY TSTF September 13, 2024 TSTF-24-07 PROJ0753 Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Attention: Rulemakings and Adjudications Staff

SUBJECT:

Petition to Amend 10 CFR 50.36, "Technical specifications" The Technical Specifications Task Force (TSTF), the Pressurized Water Reactor Owners Group (PWROG), and the Boiling Water Reactor Owners' Group (BWROG), respectfully submit the enclosed petition to amend 10 CFR 50.36, "Technical specifications." The purpose of the petition is to revise 10 CFR 50.36 to apply risk insights to the selection of technical specifications Limiting Conditions for Operation (LCO) and to update the regulation to be consistent with the guidance in the NRC's Standard Technical Specifications.

The proposed rule change directly supports NRC's initiatives:

• The "Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy Act of 2024" (the "ADVANCE Act of 2024"), Title V, Section 507, directs the Commission to report to congress the specific improvements to nuclear reactor oversight programs that maximize efficiency, including risk-informed, performance-based procedures, and to use input from the nuclear energy industry.

• The NRC's Strategic Plan for Fiscal years 2022-2026. The first safety and security objective in the plan is to provide quality licensing and oversight of nuclear facilities and radioactive materials, and the first strategy for implementing that objective is to promote risk-informed decisionmaking. The second safety and security objective is to ensure that regulatory requirements adequately support the safe and secure use of radioactive materials, and the first strategy for implementing that objective is to further risk-inform the current regulatory framework.

The Technical Specifications are the most visible and pervasive aspect of the NRC's regulatory oversight of nuclear power plants. Using a risk-informed, performance-based process to determine which requirements warrant being included in the TS will be a powerful demonstration of risk-informed decisionmaking and risk-informing the current regulatory framework in support of the NRC's Strategic Plan.

TSTF 24-07 September 13, 2024 Page 2 Should you have any questions, please do not hesitate to contact me.

Brian D. Mann Program Manager Technical Specifications Task Force

Enclosure:

Petition for Rulemaking to Amend 10 CFR 50.36, "Technical specifications" cc:

Michelle Honcharik, Technical Specifications Branch Shivani Mehta, Technical Specifications Branch

TSTF 24-07 September 13, 2024 Page 3 bcc:

Technical Specifications Task Force Chad Holderbaum, Pressurized Water Reactor Owners Group Dustin Zavada, Boiling Water Reactors Owners' Group Brett Titus, Nuclear Energy Institute

Enclosure Page 1 Petition for Rulemaking to Amend 10 CFR 50.36, "Technical specifications" Before the UNITED STATES NUCLEAR REGULATORY COMMISSION Rockville, Maryland In the Matter of a Proposed Rulemaking to Amend 10 CFR 50.36, "Technical specifications" Petitioners Name:

Technical Specifications Task Force (lead petitioner);

Pressurized Water Reactor Owners Group (co-petitioner);

Boiling Water Reactor Owners Group (co-petitioner)

Petitioners Contact Information Name:

Brian D. Mann Telephone:

804-339-7034 Address:

Excel Services Corporation, 11921 Rockville Pike, Suite 100, Rockville, MD 20852 Email:

brian.mann@excelservices.com Petition for Rulemaking by the Technical Specifications Task Force Pursuant to 10 CFR 2.802, the Technical Specifications Task Force (TSTF), the Pressurized Water Reactor Owners Group (PWROG), and the Boiling Water Reactor Owners Group (BWROG) (collectively, the petitioners) submit this petition for rulemaking requesting that the U.S. Nuclear Regulatory Commission (NRC), following public notice and opportunity for comment, amend 10 CFR 50.36, "Technical specifications." The petitioners request that the NRC promptly begin rulemaking to amend 10 CFR 50.36 to apply risk insights to the selection of technical specifications Limiting Conditions for Operation (LCO) and to update the regulation to be consistent with the guidance in the NRC's Standard Technical Specifications.

I.

Statement of Petitioners Interest The TSTF is a joint activity of the PWROG and BWROG and for over 25 years has served as the commercial nuclear power industrys subject matter experts on technical specifications (TS), and acts as the industrys point-of-contact with the NRC on generic TS-related issues. The PWROG and BWROG electric utility members operate every commercial nuclear power plant in the United States. The TSTF and its predecessor organizations represented the nuclear industry during the development of the last major change to 10 CFR 50.36.

The TSTF has developed hundreds of proposed improvements to the NRCs standard technical specifications, called "TSTF travelers." Adoption of NRC-approved TSTF travelers into

Enclosure Page 2 operating plant TS is the single largest category of license amendment requests submitted to the NRC. The TSTF has been instrumental in the development of risk-informed improvements to the TS. Should the rule be amended as requested, the TSTF will work with the PWROG and BWROG members to create TSTF travelers and industry guidance documents to implement the revised regulation.

The purpose of the petition is to request that the NRC revise 10 CFR 50.36 to apply risk insights to the selection of TS LCOs and to update the regulation to be consistent with the guidance in the NRC's Standard Technical Specifications.

II. Background and Problem Statement A.

The Regulatory History of 10 CFR 50.36, "Technical specifications" Section 182, License Applications, of the Atomic Energy Act of 1954, states (emphasis added):

a. Each application for a license hereunder shall be in writing and shall specifically state such information as the Commission, by rule or regulation, may determine to be necessary to decide such of the technical and financial qualifications of the applicant, the character of the applicant, the citizenship of the applicant, or any other qualifications of the applicant as the Commission may deem appropriate for the license. In connection with applications for licenses to operate production or utilization facilities, the applicant shall state such technical specifications, including information of the amount, kind, and source of special nuclear material required, the place of the use, the specific characteristics of the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the utilization or production of special nuclear material will be in accord with the common defense and security and will provide adequate protection to the health and safety of the public. Such technical specifications shall be a part of any license issued.

The first commercial nuclear power plant, licensed by the Atomic Energy Commission (AEC) in November 1959, was Dresden Unit 1. The initial license included a 19 page "Appendix A" (the term "Technical Specifications" does not appear in the document). The introduction stated, "The following are the principal design and performance specifications and operating limits and procedures of the Dresden Nuclear Power Station pertaining to safety." The document described items such as:

• The reactor vessel,

• The fuel assemblies,

• The fuel pellet dimensions,

• The control rod and control rod drive system,

• The steam supply system and condenser descriptions,

• Radioactive waste disposal,

• The safety system selector switch,

• Instrumentation channels and their scram trip setting, and

• The reactor enclosure.

Enclosure Page 3 The appendix also contained a list of "operating principles," but these principles did not restrict plant operation.

In 1962 the AEC approved a revised 10 CFR 50.36, "Designation of Technical Specifications,"

and added a new regulation, 10 CFR 50.59, "Authorization of changes, tests and experiments."

The purpose of the 1962 rule change was to designate certain significant design and operating limitations and procedures as technical specifications that must be adhered to in the absence of specific authorization from the Commission, and to allow other information to be revised by licensees following the guidance of 10 CFR 50.59. The rule change also added Appendix A, "Contents of Technical Specifications." Appendix A contained an extensive set of parameters to be included in the TS that continued the practice of the technical specifications being a list of design specifications.

In 1968, the AEC issued a revised 10 CFR 50.36, renamed "Technical specifications." As stated in a press release, "the Commission proposes to adopt a new system of technical specifications which will further emphasize and direct the attention of both licensee management and the AEC to those features and characteristics important to safety. The proposed system will focus on two general classes of technical matters: (1) those related to prevention of accidents; and (2) those related to limiting the consequences of accidents."

The revised regulation removed Appendix A and instead required the TS to include items in the following categories:

• Safety limits and limiting safety system settings;

• Limiting conditions for operation;

• Surveillance requirements;

• Design features; and

• Administrative controls This structure of the TS has been mostly unchanged in the intervening five decades.

In 1995, the NRC revised 10 CFR 50.36 (60 FR 36953) based on the Commission policy statement on TS improvements for nuclear power reactors (58 FR 39132). The rule change "establishes a specific set of objective criteria as guidance for determining which regulatory requirements and operating restrictions should be included in Technical Specifications."

The Policy Statement and the Statements of Consideration for the rule change provided this problem statement:

Technical Specifications cannot be changed by licensees without prior NRC approval.

However, since 1969, there has been a trend towards including in Technical Specifications not only those requirements derived from the analyses and evaluation included in the safety analysis report but also essentially all other Commission requirements governing the operation of nuclear power reactors. This extensive use of Technical Specifications is due in part to a lack of well-defined criteria (in either the body of the rule or in some other regulatory document) for what should be included in Technical Specifications. This has contributed to the volume of Technical Specifications

Enclosure Page 4 and to the several-fold increase, since 1969, in the number of license amendment applications to effect changes to the Technical Specifications. It has diverted both staff and licensee attention from the more important requirements in these documents to the extent that it has resulted in an adverse but unquantifiable impact on safety.

As discussed in the next section of this petition, the 1995 rule change added four criteria to 10 CFR 50.36(c)(2)(ii), "Limiting conditions for operation," to limit the requirements to be included in the TS. The NRC also worked with the BWROG and the organizations that became the PWROG to develop improved Standard Technical Specifications (STS)1 that clarified and reduced the level of detail in the TS. Application of the LCO criteria resulted in moving roughly one-third of the TS LCOs to licensee control. Approximately 82% of the operating nuclear power plants have adopted TS based on the STS, and improvements to the TS of all operating plants have been made based on the revised regulation.

In summary, 10 CFR 50.36 has been revised over the past sixty years to increasingly focus the TS on safety-significant requirements as knowledge and experience was gained as follows:

• The 1962 rule change designated certain significant design and operating limitations and procedures as TS.

• The 1968 rule change emphasized those features and characteristics important to safety, focused on matters related to prevention of accidents and those related to limiting the consequences of accidents.

• The 1995 rule change applied criteria to the selection of LCOs based primarily on accident analysis assumptions related to initial conditions and the assumed mitigation capabilities of analyzed accidents.

The proposed rule change continues this evolution by incorporating the consideration of plant risk into the selection of TS requirements.

B.

The Need to Risk-Inform the LCO Selection Criteria In 1995, the NRC published a policy statement titled, "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities" (60 FR 42622). In response, the NRC and the industry developed a number of risk-informed initiatives to improve the TS. The TSTF developed and the NRC approved a plant for eighteen risk-informed TS improvements. One of those initiatives proposed a change to 10 CFR 50.36 to risk-inform the selection of LCOs, but it was not pursued because it required a rule change. At this time, the industry believes that continued risk-informed improvements to the TS necessitate changes to the regulations and that other alternatives to risk-inform the TS have been pursued to a significant extent. Future 1 NUREG-1430 provides the STS for the Babcock & Wilcox plant designs.

NUREG-1431 provides the STS for the Westinghouse plant designs.

NUREG-1432 provides the STS for the Combustion Engineering plant designs.

NUREG-1433 provides the STS for the BWR/4 plant designs.

NUREG-1434 provides the STS for the BWR/6 plant designs.

NUREG-2194 provides the STS for the Westinghouse AP1000 plant design.

Enclosure Page 5 improvements to plant safety and operation can be made based on the application of risk to the TS selection criteria, which requires the proposed change to the regulation.

The 1995 change to 10 CFR 50.36 added the four criteria to 10 CFR 50.36(c)(2)(ii) to guide the selection of LCOs. Criterion 4 states, "A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety." This criteria requires the retention in the TS of LCOs that operating experience or probabilistic risk assessment have shown to be significant to public health and safety. However, there is no allowance to remove LCOs that are shown to not be significant to public health and safety.

The Commission's 1993 "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (the Final Policy Statement), states, "The purpose of Technical Specifications is to impose those conditions or limitations upon reactor operation necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety by identifying those features that are of controlling importance to safety and establishing on them certain conditions of operation which cannot be changed without prior Commission approval." Consistent with that purpose, requirements that are not of controlling importance to safety should be able to be changed using alternative regulatory methods, such as 10 CFR 50.59, without prior Commission approval.

Over sixty percent of the operating plants have incorporated a Risk Informed Completion Time Program2, which uses probabilistic risk assessment (PRA) tools to determine the appropriate time to take TS Actions when an LCO is not met (known as the Completion Time). This program has illustrated that some TS LCOs have little to no impact on plant risk and underscores the need to implement the Commissions vision to make better use of risk information in defining the TS requirements.

The current 10 CFR 50.36(c)(2)(ii) Criteria 2 and 3 require including all TS features that satisfy the criteria regardless of their risk significance. This results in low-risk systems being given an unnecessary priority in maintenance and testing. It can also result in a plant shutdown for conditions that are low risk. This needlessly diverts operator and NRC attention from higher-risk systems and activities. As the Commission stated in their policy statement on TS, the inclusion of unnecessary requirements in the TS can divert "both staff and licensee attention from the more important requirements in these documents to the extent that it has resulted in an adverse but unquantifiable impact on safety."

2 TSTF-505-A, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," approved November 21, 2018 (NRC Agencywide Documents Access and Management System (ADAMS) Accession No. ML18269A041).

Enclosure Page 6 C.

The Need to Update the Regulation to be Consistent with the Standard Technical Specifications The petitioners also propose updates to 10 CFR 50.36 to be consistent with the Commissions STS, and to remove information that is no longer applicable. These changes are proposed on a "not-to-interfere" basis with the primary change to risk-inform the TS LCO selection criteria.

• 10 CFR 50.36(a)(1) states, "A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications." Contrary to this statement, the NRC's STS (beginning with their initial publication in 1975) and all plant TS approved by the NRC do not include bases for the Design Features section of TS. Such bases are not needed, as the reasons for the Design Features requirements are clear. The regulation should be updated to reflect the Commissions practice of over fifty years.

• 10 CFR 50.36(c)(1)(ii)(A) requires the TS to include Limiting Safety System Settings (LSSS). However, the Commission's STS, and the plant TS based on the STS, do not include a TS section on LSSS. During the development of the STS, the requirements that were in a section titled, "Limiting Safety System Settings," were merged with the LCO requirements on instrumentation. Furthermore, after many years of discussion, the industry and the NRC have been unable to agree on what constitutes an LSSS. The requirement to have LSSS is inconsistent with the STS and causes confusion in the application of the regulation. Therefore, it is proposed to eliminate the requirement from the regulation.

• 10 CFR 50.36(c)(3) states that "Surveillance requirements [SRs] are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." However, it is inaccurate to state that SRs assure the quality of systems and components is maintained. There are many aspects of system or component quality that are not monitored by TS SRs and are instead verified by other means such as the American Society of Mechanical Engineers (ASME) Sections III, XI, and the Operations and Maintenance (OM) Code, vendor quality assurance audits and licensee receipt inspections. In addition, the licensee's NRC-approved Quality Assurance Program describes design control, procurement document control, control of purchased material, and inspection to ensure that the necessary quality of systems and components is maintained. None of the actions to assure quality are confirmed by TS SRs. Therefore, to make the regulation consistent with the NRC's STS, and to recognize the role of the 10 CFR 50, Appendix B, Quality Assurance Program and other regulatory requirements, the reference to confirming the quality of systems should be removed.

• 10 CFR 50.36(c)(5) states, "Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner." However, in 1993 the NRC removed recordkeeping, and review and audit requirements from the STS

Enclosure Page 7 Administrative Controls3. The STS Administrative Controls also include programs that are not described in the regulation. The description of Administrative Controls should be updated to reflect the NRC's STS.

• 10 CFR 50.36(d)(1) states, "This section shall not be deemed to modify the technical specifications included in any license issued prior to January 16, 1969. A license in which technical specifications have not been designated shall be deemed to include the entire safety analysis report as technical specifications." 10 CFR 50.36(d)(2) states, "An applicant for a license authorizing operation of a production or utilization facility to whom a construction permit has been issued prior to January 16, 1969, may submit technical specifications in accordance with this section, or in accordance with the requirements of this part in effect prior to January 16, 1969." These paragraphs are no longer applicable as all operating plants have TS included in the license. This provision may be removed to simplify the regulations.

III. Proposed Solution to the Problems Identified in the Petition Consistent with the requirements of 10 CFR 2.802(c)(1)(v), the petitioners propose the following:

• Revise the selection criteria for LCOs to state that items that satisfy 10 CFR 50.36(c)(2)(ii)(B) and (C) (that is, Criteria 2 and 3) must also be shown by probabilistic risk assessment to be significant to public health and safety.

• Update the description of 10 CFR 50.36(c)(2)(ii)(D) (that is, Criterion 4) to eliminate the reference to operating experience and to distinguish Criterion 4 from the revised Criteria 2 and 3.

• Revise the regulation to not require TS bases for the Design Features.

• Eliminate the requirement that the TS include requirements designated as LSSS.

• Revise the description of Surveillance Requirements to no longer state that SRs assure that the necessary quality of systems and components is maintained.

• Revise the Administrative Controls description to remove the requirement to include recordkeeping, and review and audit, and to include a requirement for programs.

• Eliminate the unnecessary exceptions for plants licensed prior to 1969.

3 Letter from William T. Russell (NRC) to Owners Group Chairpersons, "Content of Standard Technical Specifications, Section 5.0, 'Administrative Controls'," dated October 25, 1993.

Enclosure Page 8 IV. Proposed Changes to 10 CFR 50.36 The petitioners propose that the NRC amend 10 CFR 50.36 as follows. Removed words are struck-through, and added words are underlined. Changes are indicated by revisions bars in the right margin.

§ 50.36 Technical specifications.

(a) (1) Each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications, other than those covering design features and administrative controls, shall also be included in the application, but shall not become part of the technical specifications.

(2) Each applicant for a design certification or manufacturing license under part 52 of this chapter shall include in its application proposed generic technical specifications in accordance with the requirements of this section for the portion of the plant that is within the scope of the design certification or manufacturing license application.

(b) Each license authorizing operation of a production or utilization facility of a type described in § 50.21 or § 50.22 will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to § 50.34. The Commission may include such additional technical specifications as the Commission finds appropriate.

(c) Technical specifications will include items in the following categories:

(1) Safety limits, limiting safety system settings, and limiting control settings.

(i)(A) Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. Operation must not be resumed until authorized by the Commission. The licensee shall retain the record of the results of each review until the Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b) or § 50.22 of this part. For these reactors, the licensee shall notify the Commission as required by § 50.72 and submit a Licensee Event Report to the Commission as required by § 50.73. Licensees in these cases shall retain the records of the review for a period of three years following issuance of a Licensee Event Report.

(B) Safety limits for fuel reprocessing plants are those bounds within which the process variables must be maintained for adequate control of the operation and that must not be exceeded in order to protect the integrity of the physical system that is designed to guard

Enclosure Page 9 against the uncontrolled release or radioactivity. If any safety limit for a fuel reprocessing plant is exceeded, corrective action must be taken as stated in the technical specification or the affected part of the process, or the entire process if required, must be shut down, unless this action would further reduce the margin of safety. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. If a portion of the process or the entire process has been shutdown, operation must not be resumed until authorized by the Commission. The licensee shall retain the record of the results of each review until the Commission terminates the license for the plant.

(ii)(A) Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions. Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. If, during operation, it is determined that the automatic safety system does not function as required, the licensee shall take appropriate action, which may include shutting down the reactor. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence.

The licensee shall retain the record of the results of each review until the Commission terminates the license for the reactor except for nuclear power reactors licensed under § 50.21(b) or § 50.22 of this part. For these reactors, the licensee shall notify the Commission as required by § 50.72 and submit a Licensee Event Report to the Commission as required by § 50.73. Licensees in these cases shall retain the records of the review for a period of three years following issuance of a Licensee Event Report.

(B) Limiting control settings for fuel reprocessing plants are settings for automatic alarm or protective devices related to those variables having significant safety functions. Where a limiting control setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that protective action, either automatic or manual, will correct the abnormal situation before a safety limit is exceeded. If, during operation, the automatic alarm or protective devices do not function as required, the licensee shall take appropriate action to maintain the variables within the limiting control-setting values and to repair promptly the automatic devices or to shut down the affected part of the process and, if required, to shut down the entire process for repair of automatic devices.

The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. The licensee shall retain the record of the results of each review until the Commission terminates the license for the plant.

(2) Limiting conditions for operation. (i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met. When a limiting condition for operation of any process step in the system of a fuel reprocessing plant is not met, the licensee shall shut down that part of the operation or follow any remedial action

Enclosure Page 10 permitted by the technical specifications until the condition can be met. In the case of a nuclear reactor not licensed under § 50.21(b) or § 50.22 of this part or fuel reprocessing plant, the licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. The licensee shall retain the record of the results of each review until the Commission terminates the license for the nuclear reactor or the fuel reprocessing plant. In the case of nuclear power reactors licensed under § 50.21(b) or § 50.22, the licensee shall notify the Commission if required by § 50.72 and shall submit a Licensee Event Report to the Commission as required by § 50.73. In this case, licensees shall retain records associated with preparation of a Licensee Event Report for a period of three years following issuance of the report. For events which do not require a Licensee Event Report, the licensee shall retain each record as required by the technical specifications.

(ii) A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier, and that probabilistic risk assessment has shown to be significant to public health and safety.

(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier, and that probabilistic risk assessment has shown to be significant to public health and safety.

(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety and does not satisfy the criteria described in paragraphs (c)(2)(ii) (B) and (C) of this section.

(iii) A licensee is not required to propose to modify technical specifications that are included in any license issued before August 18, 1995 [Revised Rule Issuance Date], to satisfy the criteria in paragraph (c)(2)(ii) of this section.

(3) Surveillance requirements. Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

(4) Design features. Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c) (1), (2), and (3) of this section.

Enclosure Page 11 (5) Administrative controls. Administrative controls are the provisions relating to organization and management, procedures, programs, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in § 50.4.

(6) Decommissioning. This paragraph applies only to nuclear power reactor facilities that have submitted the certifications required by § 50.82(a)(1) and to non-power reactor facilities which are not authorized to operate. Technical specifications involving safety limits, limiting safety system settings, and limiting control system settings; limiting conditions for operation; surveillance requirements; design features; and administrative controls will be developed on a case-by-case basis.

(7) Initial notification. Reports made to the Commission by licensees in response to the requirements of this section must be made as follows:

(i) Licensees that have an installed Emergency Notification System shall make the initial notification to the NRC Operations Center in accordance with § 50.72 of this part.

(ii) All other licensees shall make the initial notification by telephone to the Administrator of the appropriate NRC Regional Office listed in appendix D, part 20, of this chapter.

(8) Written Reports. Licensees for nuclear power reactors licensed under § 50.21(b) and § 50.22 of this part shall submit written reports to the Commission in accordance with

§ 50.73 of this part for events described in paragraphs (c)(1) and (c)(2) of this section.

For all licensees, the Commission may require Special Reports as appropriate.

(d)(1) This section shall not be deemed to modify the technical specifications included in any license issued prior to January 16, 1969. A license in which technical specifications have not been designated shall be deemed to include the entire safety analysis report as technical specifications.

(2) An applicant for a license authorizing operation of a production or utilization facility to whom a construction permit has been issued prior to January 16, 1969, may submit technical specifications in accordance with this section, or in accordance with the requirements of this part in effect prior to January 16, 1969.

(3) At the initiative of the Commission or the licensee, any license may be amended to include technical specifications of the scope and content which would be required if a new license were being issued.

(e) The provisions of this section apply to each nuclear reactor licensee whose authority to operate the reactor has been removed by license amendment, order, or regulation.

Enclosure Page 12 V. Basis for Proposed Changes to 10 CFR 50.36 A.

The Need to Risk-Inform the LCO Selection Criteria The proposed change supports the NRC's goals. The Commission's Final Policy Statement states, "as a part of the Commission's ongoing program of improving Technical Specifications, it will continue to consider methods to make better use of risk and reliability information for defining future generic Technical Specification requirements." The proposed change also supports the Commission's Strategic Plan for 2022-20264. Strategy 1.1.1, is to, "Promote risk-informed decisionmaking to result in effective and efficient oversight, rulemaking, and licensing and certification activities." Strategy 1.2.1 is to, "Maintain and further risk-inform the current regulatory framework using information gained from operating experience, lessons learned, external and internal assessments, technology advances, research activities, and changes in the threat environment." The TS are the principle regulatory framework for ensuring safe plant operation. Applying risk insights to the selection of LCOs will be a significant enhancement to safe and efficient operation.

In 1995, the NRC and the industry developed a set of risk-informed initiatives to improve the TS.

The TSTF developed and the NRC approved a plan for eighteen risk-informed TS travelers. One of the proposed initiatives was to change 10 CFR 50.36 to risk-inform the selection of LCOs, but it was not pursued at the time since it required a rule change, and other improvements could be made without the time required to revise the regulation. However, the industry believes that the alternatives to revising the regulation have been largely implemented and further improvements to plant safety and operation can be made by the application of risk insights to the TS selection criteria.

When the 1995 rule change was promulgated, risk assessment tools were limited, and there were no standards for plant-specific PRA models. Now all operating plants have plant-specific PRA models that have been reviewed by the NRC against Regulatory Guide 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities." All licensees are committed to maintaining a PRA that reflects the as-built, as-operated plant, which includes incorporation of operating experience regarding human and equipment performance. As stated in Regulatory Guide 1.200, Revision 3, "The licensee is to demonstrate that the PRA model represents the current plant design and configuration and represents current operating practices and operating experience to the extent required to support the application. This demonstration can be achieved through a PRA configuration control plan that includes provisions for updating the model periodically to reflect changes that impact the significant accident sequences." The NRC's review of plant PRAs for specific applications, such as the Surveillance Frequency Control Program, includes evaluation of the licensee's processes for maintaining the PRA to reflect the as-built and as-operated plant, including operating experience. In addition, all plants have implemented risk assessment tools to satisfy 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," (the Maintenance Rule), and have adopted risk-informed TS provisions.

4 NUREG-1614, "Strategic Plan," Volume 8.

Enclosure Page 13 The proposed change revises 10 CFR 50.36(c)(2)(ii)(B) and (C) (Criterion 2 and 3) to add the condition, "and probabilistic risk assessment has shown to be significant to public health and safety." This phrase currently appears in Criterion 4. The NRC has issued 50 license amendment requests, revising the TS of 77 operating units, to convert plant TS to the format of the Standard Technical Specifications. In each of those amendments, licensees relocated many LCOs (typically about one-third of the pre-conversion LCOs) that were deemed to not satisfy any of the criteria in 10 CFR 50.36(c)(2)(ii). For each relocated LCO, the licensee was required to demonstrate, and the NRC to concur, that the LCO did not satisfy Criterion 4. Therefore, the industry and the NRC have extensive experience in applying the concept of "probabilistic risk assessment has shown to be significant to public health and safety." Application of this concept to Criterion 2 and 3 would not be different from that experience.

It is proposed that the 10 CFR 50.36(c)(2)(ii)(D) be revised to remove the separate requirement to consider operating experience in addition to PRA. As discussed above, licensees are committed to maintaining the PRA, including the consideration of operating experience.

Therefore, it is not necessary to require consideration of operating experience separately.

It is proposed that the 10 CFR 50.36(c)(2)(ii)(D) be revised to state that the items that satisfy Criterion 4 do not also satisfy Criterion 2 or 3. This distinguishes the purpose of Criteria 4 from the revised Criterion 2 and 3. Without the exclusion, all items that satisfy the revised Criterion 2 and 3 would also satisfy Criterion 4, which could be confusing and could necessitate revising the TS Bases discussion of the applicable criteria for most LCOs.

Relocating TS requirements that are not risk-significant to licensee control improves plant safety.

As stated in the Final Policy Statement, "The purpose of Technical Specifications is to impose those conditions or limitations upon reactor operation necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety by identifying those features that are of controlling importance to safety and establishing on them certain conditions of operation which cannot be changed without prior Commission approval."

PRA is an effective tool in identifying limitations on reactor operation that are of controlling importance to safety, and licensee's plant-specific PRAs have identified systems, structures, and components that have minimal impact on plant risk but have been included in TSs based on the requirements of Criterion 2 or 3. That is not to imply that those SSCs are not needed, but that the SSCs are not of controlling importance to safety. Consistent with the Commission's Final Policy Statement, such SSCs do not warrant requirements that cannot be changed without prior Commission approval.

The Commission made a similar determination when the four criteria were established in 1995.

Licensees were permitted to submit license amendment requests to remove LCO requirements that did not satisfy the four criteria. Thirty years of operating experience with relocated items has demonstrated that these relocations are not detrimental to plant safety or regulatory compliance. Approximately 80% of the operating plants have adopted TS based on the STS since the early 1990's, and in that period plant capacity factors have increased, while reactor trips and NRC violations have substantially decreased. These changes are at least partially attributable to the focus of TS on important issues by application of the criteria. The widespread use of risk tools provides an opportunity to further refine the attention of the industry and the NRC on risk-significant issues, consistent with the Commission's policy.

Enclosure Page 14 The proposed change does not mandate or approve any changes to a plants TS. This is made clear by updating the date in Paragraph (c)(2)(iii) to the issuance date of the revised regulation.

The proposed rule change provides the opportunity for licensees to apply risk insights to the selection of TS LCOs. If it is determined that an existing LCO does not satisfy the revised criteria, a licensee may submit a license amendment request for NRC review to relocate the LCO to licensee control. Relocation of an LCO to licensee control does not eliminate the requirement but provides greater flexibility in revising the requirement if desired. Any proposed change to the relocated requirement would be required to be evaluated in accordance with 10 CFR 50.59 to determine if prior NRC approval is required.

In conjunction with the rule change, the Commission's final policy statement should be updated to describe the revised Criterion 2, 3, and 4. Attachment 1 describes suggested revisions to the Commission's "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" to reflect the proposed rule change.

B.

The Need to Update the Regulation to be Consistent with the Standard Technical Specifications If 10 CFR 50.36 is revised, there are other improvements and corrections that the petitioners recommend be included in the rule change on a "not-to-interfere" basis with the primary change to risk-inform the TS LCO selection criteria.

1.

Correct the Bases Description 10 CFR 50.36(a)(1) states, "A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications." Contrary to this guidance, the NRC's STS as early as 1975 and all plant TS approved by the NRC do not include bases for the Design Features section of TS. Such bases are not needed, as the reasons for the Design Features requirements are clear.

To eliminate this inconsistency, the petitioners recommend revising 10 CFR 50.36(a)(1) to state, "A summary statement of the bases or reasons for such specifications, other than those covering design features and administrative controls, shall also be included in the application, but shall not become part of the technical specifications."

2.

Recognize that the Limiting Safety System Settings (LSSS) have been Incorporated into the LCO Requirements 10 CFR 50.36(c)(1)(ii)(A), states:

Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions. Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. If, during operation, it is determined that the automatic safety system does not function as required, the licensee shall take appropriate action, which may include shutting down the reactor. The licensee shall notify the

Enclosure Page 15 Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence.

The licensee shall retain the record of the results of each review until the Commission terminates the license for the reactor except for nuclear power reactors licensed under

§ 50.21(b) or § 50.22 of this part. For these reactors, the licensee shall notify the Commission as required by § 50.72 and submit a Licensee Event Report to the Commission as required by § 50.73. Licensees in these cases shall retain the records of the review for a period of three years following issuance of a Licensee Event Report.

Early NRC standard TS and plant TS included an LSSS section of the TS. However, this section was eliminated in the NRC's 1993 STS5 and the functions and setpoints were incorporated into the reactor protection system specification.

During the development of TSTF-493, "Clarify Application of Setpoint Methodology for LSSS Functions," between 2003-2010, the industry and the NRC debated which functions were LSSS functions and which were LCO functions. No agreement was reached, and the industry and NRC adopted a solution that did not require identifying the LSSS functions. There continues to be no common understanding on which functions are LSSS functions and which are LCO functions. Given that 80% of operating plants have TS based on the STS and do not include an LSSS section in the TS, and that there is no agreement on which functions are LSSS and which are LCOs, it is recommended that 10 CFR 50.36(c)(1)(ii)(A) be removed.

This change will not affect plant operation or any regulatory requirements because the LCOs provide equivalent requirements to the LSSS. The instrumentation LCOs contain the "settings for automatic protective devices related to those variables having significant safety functions."

The LCO settings for the automatic protective devices are "chosen [such] that automatic protective action will correct the abnormal situation before a safety limit is exceeded." The TS Actions require taking the appropriate action, which may include shutting down the reactor, if the functions are not operable.

10 CFR 50.36(c)(1)(ii)(A) states that when an LSSS does not function as required at a commercial nuclear power plant, the licensee is to follow 10 CFR 50.72 and 10 CFR 50.73. The NRC's guidance on these regulations, NUREG-1022, does not include any specific requirements for reporting LSSS malfunctions versus LCO malfunctions. A review of Licensee Event Reports for the last twenty-five years found three LERs reported under 10 CFR 50.36(c)(1)(ii)(A), and those events were also reported under other 10 CFR 50.73 criteria. Therefore, removing the LSSS reporting requirement will not affect the information being provided to the NRC.

5 NUREG-1430 provides the STS for Babcock & Wilcox plant designs.

NUREG-1431 provides the STS for Westinghouse plant designs.

NUREG-1432 provides the STS for Combustion Engineering plant designs.

NUREG 1433 provides the STS for BWR/4 plant designs, but is also representative of the BWR/2, BWR/3, and in some cases, of the BWR/5 plant design.

NUREG 1434 provides the STS for BWR/6 plant designs but is also representative in some cases of the BWR/5 plant design.

NUREG-2194 provides the STS for Westinghouse AP1000 plant designs.

Enclosure Page 16 The LSSS requirements on record retention are identical to record retention requirements in paragraph 50.36(c)(2) on LCOs.

In summary, eliminating 10 CFR 50.36(c)(1)(ii)(A) will bring the regulations into alignment with the NRC's published STS and most issued plant TS, will eliminate an ambiguous requirement, and will not affect plant safety or the NRC's regulatory oversight.

3.

Revise the Surveillance Requirement Description to be Consistent with the STS 10 CFR 50.36(c)(3) states, "Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met."

The description that Surveillance Requirements assure the necessary quality of systems and components is maintained is outdated and should be revised.

In the NRC's STS, Surveillance Requirement 3.0.1 states, "Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3." Therefore, the TS unambiguously link Surveillance Requirements to ensuring that the corresponding LCOs will be met.

Safety Limits do not have specific Surveillance Requirements, but LCOs are constructed to ensure the Safety Limits are met and those LCOs have corresponding SRs. Therefore, the TS link Surveillance Requirements to the Safety Limits.

However, there are many aspects related to the quality of systems that are not the subject of Surveillance Requirements. These aspects may or may not affect meeting the associated LCOs.

As stated in the NRC's Inspection Manual Chapter (IMC) 0326, "Operability Determinations,"

"Not all SSC [structure, system or component] functions described in the CLB [current licensing basis] are specified safety functions required for operability." For example, the licensee's NRC-approved Quality Assurance Program describes design control, procurement document control, control of purchased material, and inspection to ensure that the necessary quality of systems and components is maintained. In addition, 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," requires all licensees to monitor the performance or condition of structures, systems, or components (SSCs), against licensee-established goals, in a manner sufficient to provide reasonable assurance that these SSCs are capable of fulfilling their intended functions. This regulatory requirement is not implemented by TS Surveillance Requirements.

Therefore, to make the regulation consistent with the NRC's STS and NRC-approved plant TS, and to recognize the role of the 10 CFR 50, Appendix B, Quality Assurance Program, and the 10 CFR 50.65 requirements, 10 CFR 50.36(c)(3) should be revised to eliminate the statement that Surveillance Requirements assure the necessary quality of systems and components is maintained. This change will not affect the implementation of any existing Surveillance Requirements.

Enclosure Page 17

4.

Revise Administrative Controls Description to be Consistent with the STS 10 CFR 50.36(c)(5) states, "Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner."

In the STS (adopted by 82% of licensees), most of the Administrative Controls are related to programs, which are descriptions of licensee programs that implement TS requirements, such as steam generator inspection, diesel fuel oil testing, ventilation filter testing, and control room inleakage testing. Some, but not all, of these programs are referenced from LCOs or Surveillance Requirements.

Recordkeeping and review and audit requirements were removed from the STS Administrative Controls in 19936.

To reflect these changes and to be consistent with the NRC's STS and most plant TS, 10 CFR 50.36(c)(5) should be revised to include programs in the content of Administrative Controls and to remove reference to recordkeeping, review, and audit.

5.

Delete Outdated Paragraph (d) 10 CFR 50.36(d)(1) states, "This section shall not be deemed to modify the technical specifications included in any license issued prior to January 16, 1969. A license in which technical specifications have not been designated shall be deemed to include the entire safety analysis report as technical specifications."

10 CFR 50.36(d)(2) states, "An applicant for a license authorizing operation of a production or utilization facility to whom a construction permit has been issued prior to January 16, 1969, may submit technical specifications in accordance with this section, or in accordance with the requirements of this part in effect prior to January 16, 1969."

These paragraphs are no longer applicable. All operating plants have TS included in their licenses. All future license applicants are required to propose TS by 10 CFR 50.36(a)(1) and the NRC is required by 10 CFR 50.36(b) to issue TS with the license. Therefore, these paragraphs can be removed without effect.

VI. Additional Rulemaking Information A.

Consistency with the Atomic Energy Act The proposed petition does not affect compliance with Section 182, License Applications, of the Atomic Energy Act of 1954, that requires licensees to include technical specifications.

6 Letter from William T. Russell (NRC) to Lee Bush (Westinghouse Technical Specifications Committee), et.al.,

"Content of Standard Technical Specifications, Section 5.0, 'Administrative Controls'," dated October 25, 1993.

Enclosure Page 18 B.

Priority To assist the NRC staff in prioritizing the proposed rulemaking, Attachment 2 contains a discussion of the common prioritization of rulemaking factor selection criteria.

C.

Environmental Impact Under NEPA This petition does not involve a major Federal action significantly affecting the quality of the human environment. Therefore, an environmental impact statement is not required. Granting of future requested amendments to implement the requested regulation change would not alter the environmental impact of the licensed activities described in the Final Environmental Impact Statement for each facility, as prescribed in the National Environmental Policy Act and 10 CFR Part 51. In particular, the rule change requested in this petition would not significantly increase the probability or consequences of an accident. Granting of future requested amendments to implement the requested regulation change would not involve changes in the types or quantities of radiological effluents that may be released offsite, nor a significant increase in public or occupational radiation exposure since there would be no change to facility operations that could create a new or affect a previously analyzed accident or release path. With regard to non-radiological impacts, the rule change requested in this petition involves neither changes to non-radiological plant effluents, nor changes in activities that would adversely affect the environment. Therefore, there are no significant non-radiological impacts associated with the petition.

D.

Paperwork Reduction Act Statement The proposed petition does not contain any new or amended information requirements that would be subject to the Paperwork Reduction Act.

E.

Regulatory Backfit Analysis This petition does not involve any action constituting a backfit as addressed in 10 CFR 50.109.

The proposed change permits, but does not require, a licensee to submit a license amendment request to relocate existing TS LCOs that are not risk significant to licensee control. As stated in NUREG-1409, "Backfitting Guidelines," (July 1990), Section 2.1.1, "The backfit rule applies to actions that impose positions or requirements on licensees; it does not apply to requested actions that are optional or voluntary."

Page A1-1 Suggested Revision to the Commissions, "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" to Reflect the Proposed Rule Change The petitioners proposes that the Commission revise the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," as shown below to reflect the proposed changes to the rule. As stated in the Statements of Consideration for the 1995 10 CFR 50.36 rule change, "The Commission has decided not to withdraw the final policy statement because it contains detailed discussions of the four criteria and guidance on how the NRC staff and licensees should apply the criteria." The petitioners recommend that the Commission's policy statement be retained and revised to reflect the changes to the 10 CFR 50.36 criteria, as shown below in the following excerpt from the policy statement.

Changes are indicated by revisions bars in the right margin. Removed words are struck-through, and added words are underlined.

Excerpt From the Commission's Policy Statement The following criteria delineate those constraints on design and operation of nuclear power plants that are derived from the plant safety analysis report or PSA information and that belong in Technical Specifications in accordance with 10 CFR 50.36 and the purpose of Technical Specifications stated above.

Criterion 1 Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Discussion of Criterion 1 A basic concept in the adequate protection of the public health and safety is the prevention of accidents. Instrumentation is installed to detect significant abnormal degradation of the reactor coolant pressure boundary so as to allow operator actions to either correct the condition or to shut down the plant safely, thus reducing the likelihood of a loss-of-coolant accident.

This criterion is intended to ensure that Technical Specifications control those instruments specifically installed to detect excessive reactor coolant system leakage. This criterion should not, however, be interpreted to include instrumentation to detect precursors to reactor coolant pressure boundary leakage or instrumentation to identify the source of actual leakage (e.g., loose parts monitor, seismic instrumentation, valve position indicators).

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents Page A1-2 a challenge to the integrity of a fission product barrier, and that probabilistic risk assessment has shown to be significant to public health and safety.

Discussion of Criterion 2 Another basic concept in the adequate protection of the public health and safety is that the plant shall be operated within the bounds of the initial conditions assumed in the existing Design Basis Accident and Transient analyses and that the plant will be operated to preclude unanalyzed transients and accidents. These analyses consist of postulated events, analyzed in the FSAR, for which a structure, system, or component must meet specified functional goals.

These analyses are contained in Chapters 6 and 15 of the FSAR (or equivalent chapters) and are identified as Condition n, HI, or IV events (ANSI N 18.2) (or equivalent) that either assume the failure of or present a challenge to the integrity of a fission product barrier.

As used in Criterion 2, process variables are only those parameters for which specific values or ranges of values have been chosen as reference bounds in the Design Basis Accident or Transient analyses and which are monitored and controlled during power operation such that process values remain within the analysis bounds.

Process variables captured by Criterion 2 are not, however, limited to only those directly monitored and controlled from the control room. These could also include other features or characteristics that are specifically assumed in Design Basis Accident and Transient analyses even if they cannot be directly observed in the control room (e.g., moderator temperature coefficient and hot channel factors).

The purpose of this criterion is to capture those process variables that have initial values assumed in the Design Basis Accident and Transient analyses, and which are monitored and controlled during power operation. As long as these variables are maintained within the established values, risk to the public safety is presumed to be acceptably low. This criterion also includes active design features (e.g., high pressure /low pressure system valves and interlocks) and operating restrictions (pressure/temperature limits) needed to preclude unanalyzed accidents and transients.

It is also the intent of this criterion that those process variables, design features, or operating restrictions be captured in the Technical Specifications only if probabilistic risk assessment (PRA) demonstrates the constraint as being significant to public health and safety, consistent with the Commissions Safety Goal and Severe Accident Policies.

Otherwise, the process variables, design features, or operating restrictions may be relocated to licensee control.

Criterion 3 A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes Page A1-3 the failure of or presents a challenge to the integrity of a fission product barrier, and that probabilistic risk assessment has shown to be significant to public health and safety.

Discussion of Criterion 3 A third concept in the adequate protection of the public health and safety is that in the event that a postulated Design Basis Accident or Transient should occur, structures, systems, and components are available to function or to actuate in order to mitigate the consequence of the Design Basis Accident or Transient. Safety sequence analyses or their equivalent have been performed in recent years and provide a method of presenting the plant response to an accident. These can be used to define the primary success paths.

A safety sequence analysis is a systematic examination of the actions required to mitigate the consequences of events considered in the plants Design Basis Accident and Transient analyses, as presented in Chapters 6 and 15 of the plants FSAR (or equivalent chapters).

Such a safety sequence analysis considers all applicable events, whether explicitly or implicitly presented. The primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criteria), so that the plant response to Design Basis Accidents and Transients limits the consequences of these events to within the appropriate acceptance criteria.

It is the intent of this criterion to capture into Technical Specifications only those structures, systems, and components that are part of the primary success path of a safety sequence analysis. Also captured by this criterion are those support and actuation systems that are necessary for items in the primary success path to successfully function. The primary success path for a particular mode of operation does not include backup and diverse equipment (e.g., rod withdrawal block which is a backup to the average power range monitor high flux trip in the startup mode, safety valves which are backup to low temperature overpressure relief valves during cold shutdown).

It is also the intent of this criterion that those structures, systems, and components, be captured in the Technical Specifications only if PRA demonstrates the structures, systems, or components as being significant to public health and safety, consistent with the Commissions Safety Goal and Severe Accident Policies. Otherwise, the structures, systems, or components may be relocated to licensee control.

Criterion 4 A structure, system, or component which operating experience or probabilistic risk safety assessment has shown to be significant to public health and safety and does not satisfy the criteria described in paragraphs (c)(2)(ii) (B) and (C) of this section.

Discussion of Criterion 4 It is the Commission policy that licensees retain in their Technical Specifications LCOs, action statements and Surveillance Requirements for the following systems (as Page A1-4 applicable), which operating experience and PRA PSA has have generally shown to be significant to public health and safety and that do not satisfy Criterion 2 or 3. and any other structures, systems, or components that meet this criterion:

• Reactor Core Isolation Cooling/

• Isolation Condenser,

• Residual Heat Removal,

• Standby Liquid Control, and

• Recirculation Pump Trip.

The Commission recognizes that other structures, systems, or components may meet this criterion. Plant-and design-specific PSAs have yielded valuable insight to unique plant vulnerabilities not fully recognized in the safety analysis report Design Basis Accident or Transient analyses. It is the intent of this criterion that those requirements that PRA PSA or operating experience exposes as significant to public health and safety, consistent with the Commissions Safety Goal and Severe Accident Policies, be retained or included in Technical Specifications.

The Commission expects that licensees, in preparing their Technical Specification related submittals, will utilize any plant-specific PRA PSA or risk survey and any available literature on risk insights and PRAs PSAs. This material should be employed to strengthen the technical bases for those requirements that remain in Technical Specifications, when applicable, and to verify that none of the requirements to be relocated contain constraints of prime importance in limiting the likelihood or severity of the accident sequences that are commonly found to dominate risk.

Similarly, the NRC staff will also employ risk insights and PRAs PSAs in evaluating Technical Specifications related submittals. Further, as a part of the Commission's ongoing program of improving Technical Specifications, it will continue to consider methods to make better use of risk and reliability information for defining future generic Technical Specification requirements.

Justification Terminology The Commission's policy statement used the term "probabilistic safety assessment" or the acronym "PSA." The 10 CFR 50.36 rule change updated the terminology to "probabilistic risk assessment." The corresponding acronym is "PRA." The policy statement should be updated to be consistent with the published regulation.

Page A1-5 Criteria 2 and 3 The discussions of Criteria 2 and 3 are revised to state that Technical Specifications constraints that meet those criteria should only be retained in the Technical Specifications if they are significant to public health and safety, consistent with the Commissions Safety Goal and Severe Accident Policies. This phrase, "significant to public health and safety, consistent with the Commissions Safety Goal and Severe Accident Policies," currently appears in Criterion 4.

Criterion 4 The discussion of Criterion 4 is revised to state that it includes systems that do not satisfy Criteria 2 or 3.

The Commission's policy statement included a list of four systems that the policy declared should be retained in the TS because operating experience and PRA have generally shown them to be significant to public health and safety. Those systems are:

• Reactor Core Isolation Cooling/Isolation Condenser,

• Residual Heat Removal,

• Standby Liquid Control, and

• Recirculation Pump Trip.

Extensive research by the industry and the NRC have been unable to locate any documentation or basis for the conclusion that operating experience and PRA found that the listed systems are significant to public health and safety. Modern operating experience and plant PRA evaluations do not support the conclusion that these systems are significant to public health and safety. The final policy statement should be revised to remove the list of systems. The change will permit a licensee to evaluate whether these systems satisfy Criterion 4 and request NRC approval for relocation if they do not.

Page A2-1 Discussion of the Common Prioritization of Rulemaking Factor Selection Criteria The NRC internal method described in "FY23/24 Guidance on Common Prioritization of Rulemaking Factor Selection Criteria," (ADAMS Accession No. ML23018A148) provides guidance to the NRC staff in determining the appropriate factors to be used for each rulemaking activity in the Common Prioritization of Rulemaking (CPR) process. The CPR process requires selecting values for the following four factors for each rulemaking:

• Factor A: NRC Strategic Plan strategic goals

• Factor B: NRC Principles of Good Regulation

• Factor C: Governmental priority

• Factor D: Stakeholder confidence To assist the NRC staff in the prioritization of the proposed rulemaking, the petitioners provide the following information for consideration by the NRC staff in evaluating each factor.

Factor A: NRC Strategic Plan Strategic Goals The NRCs Strategic Plan for FY 2022-2026 addresses the NRCs goal to ensure the safe and secure use of radioactive materials. The Strategic Plan contains three objectives and several strategies for each objective under this goal. The petitioners believe that the proposed rulemaking would be a significant contributor to satisfying two of the Strategic Plan objectives and strategies.

Strategic Plan Safety and Security Objective 1: Provide quality licensing and oversight of nuclear facilities and radioactive materials.

Objective 1 Strategy 1: Promote risk-informed decisionmaking to result in effective and efficient oversight, rulemaking, and licensing and certification activities.

Strategic Plan Safety and Security Objective 2: Ensure that regulatory requirements adequately support the safe and secure use of radioactive materials.

Strategy 1.2.2 - Maintain and further risk-inform the current regulatory framework using information gained from operating experience, lessons learned, external and internal assessments, technology advances, research activities, and changes in the threat environment.

The Technical Specifications are the most visible and pervasive aspect of the NRC's regulatory oversight of power plants. Using a risk-informed process to determine which requirements warrant being included in the TS will be a powerful demonstration of risk-informed decisionmaking and risk-informing the current regulatory framework in support of the NRC's Strategic Plan. Therefore, the petitioners believe the petition warrants a "high" ranking and a value of 20 points.

Page A2-2 Factor B: NRC Principles of Good Regulation The NRCs Strategic Plan for FY 2022-2026 states that it is the vision of the agency to demonstrate the principles of good regulation in implementing the NRCs safety and security mission. The petitioners believe that the proposed rulemaking would support several of the principles.

Efficiency: The American taxpayer, the rate-paying customer, and licensees are all entitled to the best possible management and administration of regulatory activitiesRegulatory activities should be consistent with the degree of risk reduction they achieve.

a. Provide the most cost-efficient methods for implementing the requirements Clarity: Regulations should be coherent, logical, and practical. There should be a clear nexus between regulations and agency goals and objectives whether explicitly or implicitly stated. Agency positions should be readily understood and easily applied.

a. Improve the clarity of agency requirements to ensure full compliance when promulgated.

b. Example: The NRC promulgates a rule that would clarify the agencys safety or security mission or requirements.

Reliability: Regulations should be based on the best available knowledge from research and operations experience. Systems interactions, technical uncertainties, and the diversity of licensees and regulatory activities must all be taken into account so that risks are maintained at an acceptably low level.

a. Ensure the reliability of the requirements with which licensees and applicants, etc.,

must be in compliance.

Using a risk-informed process to determine which regulatory requirements rise to the level of being included in the TS will increase efficiency by reserving the resource-intensive license amendment process for risk-significant requirements. Applying risk insights to the selection of TS requirements will increase regulatory clarity by applying to the TS an approach the NRC has already formulated in other risk-informed regulations, such as 10 CFR 50.69. Applying risk insights to the selection of TS requirements will improve reliability by ensuring that risk-significant systems are included in the TS. Therefore, the petitioners believe the petition warrants a "high" ranking and a value of 5 points.

Factor C: Governmental priority This weighting factor is based on two considerations. First, it can be used to account for a rulemaking initiated by a congressional mandate or other Federal requirements and thus requiring a greater agency focus (priority). Second, it can be used to adjust the relative priorities of rulemakings across the agency so that the integrated list of rulemakings appropriately reflects the agencys strategic priorities.

Page A2-3 The proposed rulemaking addresses several NRC requirements in the Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy Act, also known as the ADVANCE Act of 2024.

ADVANCE Act, Section 507, "Improving Oversight and Inspection Programs," states:

"assess specific elements of oversight and inspections that may be modified by the use of technology, improved planning, and continually updated risk-informed, performance-based assessment" "identify and assess measures to improve oversight and inspections, including-elimination of areas of duplicative or otherwise unnecessary activities;"

"increased use of inspection approaches that balance the level of resources commensurate with safety significance; "

"increased review of the use of inspection program resources based on licensee performance;"

"assess the ability of the Commission, consistent with the mission of the Commission, to enable licensee innovations that may advance nuclear reactor operational efficiency and safety" "identify specific actions that the Commission may take to incorporate into the training, inspection, oversight, and licensing activities, and regulations, of the Commission."

Applying risk-insights to the selection of TS requirements supports these mandates. Verifying compliance with TS requirements is a principal inspection activity of the NRC. Relocating low-safety-significance requirements from the TS to licensee control under 10 CFR 50.59 permits licensee innovation that balances the level of resources commensurate with the safety significance, and eliminates unnecessary activities. The petition also promotes efficient and safe plant operation by focusing licensee attention on risk-significant requirements. Therefore, the petitioners believe the petition warrants a "high" ranking and a value of 10 points.

Factor D: Stakeholder confidence The NRCs Strategic Plan for FY 2022-2026 includes numerous strategies under the two Stakeholder Confidence Objectives.

Stakeholder Confidence Objective 1: Engage stakeholders in NRC activities in an effective and transparent manner.

Strategy 1. Foster proactive and meaningful interactions with States, Tribes, other governmental and nongovernmental organizations, the regulatory industry, the international regulatory community, and other members of the public. (emphasis added,

p. 6 of Prioritization Guidance)

Page A2-4 Strategy 2. Provide a fair and timely process to allow public involvement in NRC decisionmaking. (p. 6 of Prioritization Guidance)

Stakeholder Confidence Objective 2: Uphold an NRC decisionmaking process that is data driven and evidence based while ensuring information is available and accessible to interested stakeholders.

Strategy 2. Develop effective communication strategies to explain how risk and uncertainty are addressed and considered in the decisionmaking process. (p. 6 of Prioritization Guidance)

The prioritization guidance states that Factor D is a weighting factor for a rule considered to be of significant interest to industry, nongovernmental organizations, or other members of the public. This factor could consider regulatory incentives such as eliminating the need for license amendments or reducing regulatory burden.

The proposed rulemaking is of high interest to the nuclear industry, as demonstrated by it being petitioned by the two industry Owners Groups which represent all operating U.S. plants. It will result in significant burden reduction as some TS requirements which can currently only be changed with prior Commission approval may be revised by licensees under the requirements of 10 CFR 50.59. The TS are the most visible and pervasive aspect of the NRC's regulatory oversight of power plants and moving low-safety-significant requirements from the TS to licensee control will result in a substantial burden reduction. Changes to plant-specific TS to implement the proposed change in the regulation will utilize the 10 CFR 50.90 license amendment process, which is a fair and timely process that permits public involvement. Such license amendments will be based on data and evidence gathered by licensees and presented in an available and accessible manner to interested stakeholders. The licensee's amendment requests and NRC's Safety Evaluations are effective communication methods that will describe how risk and uncertainty are addressed. The petitioners believe that the proposed rulemaking would be a significant contributor toward one or more considerations and warrants a "high" ranking and a value of 10 points.

Conclusion In conclusion, the petitioners believe that the proposed rulemaking warrants a "High" priority with a total prioritization score of 45 points.