NL-25-0231, License Amendment Request Proposing Removal of RCS Subcooling Margin Monitor from Technical Specifications List of Post Accident Monitoring Instrumentation
| ML25181A816 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/29/2025 |
| From: | Coleman J Southern Nuclear Operating Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| NL-25-0231 | |
| Download: ML25181A816 (1) | |
Text
Regulatory Affairs 3535 Colonnade Parkway Birmingham, AL 35243 205 992 5000 June 30, 2025 Docket Nos.: 50-348 NL-25-0231 50-364 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Units 1 & 2 License Amendment Request Proposing Removal of RCS Subcooling Margin Monitor from Technical Specifications List of Post Accident Monitoring Instrumentation Ladies and Gentlemen:
Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) requests a license amendment to the Joseph M. Farley Nuclear Plant (FNP) Units 1 and 2 Renewed Facility Operating Licenses NPF-2 and NPF-8, respectively. The proposed amendment would revise Technical Specifications (TS) Table 3.3.3-1, Post Accident Monitoring Instrumentation to delete Function 10, RCS Subcooling Margin Monitor. This proposed change is being made in accordance with WCAP-15981-NP-A (Non-Proprietary), Rev. 0, Post Accident Monitoring Instrumentation Re-definition for Westinghouse NSSS Plants.
WCAP-15981-NP-A provides technical justification for identifying the post accident monitoring (PAM) instrumentation that should be included in the TS for Westinghouse NSSS plants.
The enclosure provides a basis for the proposed change. Attachment 1 contains marked-up TS pages. Attachment 2 contains revised (clean) TS pages. Attachment 3 provides the marked-up TS Bases pages for information only. Attachment 4 provides the marked-up Updated Final Safety Analysis Report pages for information only.
SNC requests approval of the proposed amendment within 12 months of completion of the NRCs acceptance review. Once approved, the amendment shall be implemented within 60 days.
In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Alabama State Official.
This letter contains no NRC commitments. If you have any questions, please contact Ryan Joyce at 205.992.6468.
U. S. Nuclear Regulatory Commission NL-25-0231 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on the
___ day of ________ 2025.
Respectfully submitted, Jamie M. Coleman Director, Regulatory Affairs Southern Nuclear Operating Company JMC/was/cbg
Enclosure:
Basis for Proposed Changes Attachments: 1. Proposed Technical Specification Changes (Marked-up Pages)
- 2. Revised Technical Specification Pages (Clean Typed)
- 3. Proposed Technical Specification Bases Pages (Marked-up) - For Information Only
- 4. Proposed Updated Final Safety Analysis Report Pages (Marked-up) -
For Information Only cc:
Regional Administrator, Region ll NRR Project Manager - Farley 1 & 2 Senior Resident Inspector - Farley 1 & 2 Director, Alabama Office of Radiation Control RType: CFA04.054
Joseph M. Farley Nuclear Plant - Units 1 & 2 License Amendment Request Proposing Removal of RCS Subcooling Margin Monitor from Technical Specifications List of Post Accident Monitoring Instrumentation Enclosure Basis for Proposed Changes
Enclosure to NL-25-0231 Basis for Proposed Changes E-1 1.0
SUMMARY
DESCRIPTION In accordance with 10 CFR 50.90, Application for amendment of license, construction permit or early site permit, Southern Nuclear Operating Company (SNC) requests a license amendment to the Joseph M. Farley Nuclear Plant (FNP) Units 1 and 2 Renewed Facility Operating Licenses NPF-2 and NPF-8, respectively. The proposed amendment revises Technical Specifications (TS) Table 3.3.3-1, Post Accident Monitoring Instrumentation to delete Function 10, Reactor Coolant System (RCS) Subcooling Margin Monitor (SMM).
2.0 DETAILED DESCRIPTION 2.1 Current Licensing Basis This license amendment request seeks to amend the FNP Units 1 and 2 Renewed Facility Operating Licenses NPF-2 and NPF-8. The current licensing basis in TS 3.3.3 identifies essential Post Accident Monitoring (PAM) instruments addressing the recommendations of Regulatory Guide 1.97 (Reference 1) as required by Supplement 1 to NUREG-0737 (Reference 2).
The primary purpose of the PAM instrumentation is to display unit variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Accidents (DBAs).
The instrument channels required to be operable by TS 3.3.3 include two classes of parameters identified during unit specific implementation of Regulatory Guide 1.97 as Type A and certain Category I variables. The core subcooling monitor (RCS SMM) is currently listed in the Farley Updated Final Safety Analysis Report (UFSAR) Table 7.5-1, (Sheet 1 of 16) Post Accident Instrumentation as a Type A, Category 2 variable, and it is also listed on UFSAR Table 7.5-1, (Sheet 3 of 16) as a Type B, Category 2 variable. Type A variables provide primary information required for the control room operator to take specific manually controlled actions for which no automatic control is provided and are required for safety systems to accomplish their safety functions for DBAs. The operability of the PAM instrumentation ensures there is sufficient information available on selected unit parameters to monitor and assess unit status following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97. TS Table 3.3.3-1 lists all Type A and certain Category I variables identified by the unit specific Regulatory Guide 1.97 analyses, as amended by the NRC's SER (Reference 3).
RCS Subcooling Margin is provided to determine safety injection termination and reinitiation and depressurization and cooldown progression. The RCS SMM measures saturation/superheat margin. The function of the RCS SMM is to calculate the subcooled margin which is the difference between the measured temperature of the reactor coolant and the saturation temperature. The saturation temperature is calculated from the minimum primary system pressure input. A maximum or representative temperature input is used for the measured value, which could come from a Resistance Temperature Detector (RTD) loop, or a representative core exit thermocouple (CET).
Enclosure to NL-25-0231 Basis for Proposed Changes E-2 2.2 Proposed Licensing Basis Change The proposed change is to remove the RCS SMM Function from TS 3.3.3, Post Accident Monitoring (PAM) Instrumentation by deleting Function 10 in TS Table 3.3.3-1, Post Accident Monitoring Instrumentation. The RCS SMM indication will be retained in Farley UFSAR Table 7.5-1 as a Type B, Category 2 variable as shown in Attachment 4 to this enclosure.
The proposed Technical Specification change revises the Regulatory Guide 1.97 (Reference 1) instrumentation contained in TS 3.3.3 to be consistent with the technical basis for accident monitoring instrumentation identified in WCAP-15981-NP-A, "Post Accident Monitoring Instrumentation Re-Definition for Westinghouse NSSS Plants" (Reference 4). This change includes evaluating the current Regulatory Guide 1.97 categorization of the affected instrumentation with respect to its function as a post accident monitoring instrument based on WCAP-15981-NP-A. The results of the WCAP-15981-NP-A evaluation provided in this enclosure is for the sole purpose of determining whether it is appropriate to retain the RCS SMM in TS 3.3.3, Post Accident Monitoring (PAM) Instrumentation. The current FNP response to Regulatory Guide 1.97 will not be changed as a result of this proposed change. Therefore, there are no changes to the FNP response to Regulatory Guide 1.97 or the plant design associated with this amendment request.
The marked-up TS pages relating to this change are provided in Attachment 1. In addition, draft clean TS typed pages are provided in Attachment 2. The changes to the Technical Specifications Bases relating to this change are provided in Attachment 3 for information only.
The changes to the Updated Final Safety Analysis Report relating to this change are provided in for information only.
2.3 Reason for Proposed Change The proposed change will make the content of the FNP TS 3.3.3, "Post Accident Monitoring (PAM) Instrumentation" more consistent with the corresponding Standard Technical Specifications (STS), NUREG-1431, Revision 5, (Reference 5) 3.3.3, "Post Accident Monitoring (PAM) Instrumentation" while maintaining the FNP response to Regulatory Guide 1.97 unchanged. The proposed TS 3.3.3 change will also eliminate the need for plant shutdown as currently required by TS 3.3.3 Condition E in the event that both RCS Subcooling Margin Monitor channels are inoperable. Note that the proposed activity removes the RCS SMM from the Technical Specifications; it is not proposing to change the safety classification, operation, or maintenance of the RCS SMM System for the FNP Units. Any proposed future changes to the RCS SMM following implementation of this proposed change will be controlled by existing design control procedures and regulatory processes.
3.0 TECHNICAL EVALUATION
The FNP Regulatory Guide 1.97 Analysis and TS 3.3.3, Post Accident Monitoring (PAM)
Instrumentation PAM functions have been evaluated consistent with the guidance provided by WCAP-15981-NP-A for the proposed change. The plant-specific implementation of WCAP-15981-NP-A requires a review of:
- 1) Design Basis Accidents
Enclosure to NL-25-0231 Basis for Proposed Changes E-3
- 2) Emergency Operating Procedures
- 5) Emergency Plan A summary of these reviews is discussed in Section 3 below. Additionally, the NRC Final Safety Evaluation contained in WCAP-15981-NP-A requires that amendment requests based on WCAP-15981-NP-A submit additional plant-specific information. These requirements are addressed in Section 3.6 of this enclosure.
3.1 Design Basis Accidents The instrumentation that provides the primary information that is essential for the direct accomplishment of the specified manual actions (including long-term recovery actions) for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis accidents or transients was identified from the design basis accident analyses documented in FNP UFSAR Chapter 15.
The manual actions specifically credited in the design basis accident analyses for which no automatic control is provided are described in detail below.
Inadvertent Operation of the Emergency Core Cooling System During Power Operation (UFSAR Section 15.2.14)
In the event of an Operation of the Emergency Core Cooling System During Power Operation, the DBA analyses assume that the operators will terminate Safety Injection (SI) according to the plant Emergency Operating Procedures (EOPs). The SI termination criteria in the EOPs for this DBA, and any other event with SI operating, is based on a combination of the pressurizer level, RCS pressure, secondary heat sink, and RCS subcooling indications.
Steam Line Break (UFSAR Section 15.4.2.1)
In the event of a Steam Line Break (SLB), the DBA analyses assume that the operators will terminate SI. The primary diagnosis of a SLB condition is based on steam generator (SG) pressures. Comparison of SG steam flow between the SGs and SG water level can also be used to diagnose a SLB accident. Termination of SI is based on a combination of pressurizer level, RCS pressure, secondary heat sink, and RCS subcooling.
Steam Generator Tube Rupture (UFSAR Section 15.4.3)
The specific operator actions assumed in the DBA analysis are:
x Operator action to initiate cooldown using the intact SGs - based on ruptured SG pressure (cooldown target) and RCS pressure and temperature (maintain subcooling),
x Operator action to initiate RCS depressurization using the pressurizer spray or Power Operated Relief Valves (PORVs) - based on RCS pressure, pressurizer level, and RCS subcooling to initiate SI termination - based on SI termination criteria of RCS subcooling, secondary heat sink, RCS pressure, RCS temperature and pressurizer level.
Operator actions to cooldown and depressurize the RCS depend on several different instrumentation functions including pressurizer level and RCS subcooling indications which are used to control/terminate SI flow during the depressurization to assure that the pressurizer is not overfilled.
Enclosure to NL-25-0231 Basis for Proposed Changes E-4 3.2 Emergency Operating Procedures The FNP Emergency Operating Procedures (EOPs) provide instructions for the optimal recovery from plant events that result in either a reactor trip or a safety injection signal. While the EOPs are consistent with the plant design and licensing basis, they also provide recovery instructions for events that are outside the plant design and licensing basis. The EOPs also utilize a broad range of plant instrumentation to diagnose plant conditions and monitor the performance of equipment and systems.
The EOPs for FNP Unit 1 and Unit 2 were reviewed to identify key operator actions that would be implemented during an optimal recovery from Design Basis Accidents. The RCS SMM indication is used in multiple FNP EOPs, but if the RCS SMM indication is not available, FNP Operators are trained to use alternate indications to calculate subcooling. The proposed change to remove the RCS SMM Function from the PAM TS 3.3.3 would retain the core subcooling monitor (RCS SMM) Function in FNP UFSAR Table 7.5-1 as a Type B, Category 2 variable and would not require procedural step changes to the FNP EOPs.
3.3 Probabilistic Risk Assessment The important operator actions to prevent core damage for FNP can be derived from the Probabilistic Risk Assessment (PRA).
The FNP PRA provides an accurate representation of the design and operation of FNP Units 1 and 2. The FNP PRA models described below have been peer reviewed and there are no PRA upgrades that have not been peer reviewed.
Internal Events and Internal Flooding The SNC PRA Configuration Control processes ensure that the PRA model used in this application acceptably reflects the as-built and as-operated plant for FNP.
Fire Hazards The internal Fire PRA model was developed consistent with methods previously accepted by the NRC. The SNC PRA Configuration Control processes ensure that the PRA model used in this application acceptably reflects the as-built and as-operated plant for FNP.
Seismic Hazards FNP meets the EPRI 3002017583 (Reference 6) Tier 1 criteria for a Low Seismic Hazard/High Seismic Margin site. The Tier 1 criterion (i.e., basis) in EPRI 3002017583 is a comparison of the ground motion response spectrum (GMRS, derived from the seismic hazard) to the safe shutdown earthquake (SSE, i.e., seismic design basis capability).
Other External Hazards All external hazards, except for seismic, were screened for applicability to FNP per a plant-specific evaluation in accordance with GL 88-20 (Reference 7) and updated to use the criteria in ASME PRA Standard RA-Sa-2009.
PRA Maintenance and Updates The SNC PRA Configuration Control processes ensure that the applicable PRA models continue to reflect the as-built and as-operated plant for FNP. The process delineates the
Enclosure to NL-25-0231 Basis for Proposed Changes E-5 responsibilities and guidelines for updating the PRA models, and includes criteria for both regularly scheduled and interim PRA model updates. The process includes provisions for monitoring potential areas affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, and industry operational experience) for assessing the risk impact of unincorporated changes, and for controlling the model and associated computer files.
PRA Review Process Results The PRA models described above have been assessed against RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2 (Reference 8) and Revision 3 (Reference 13).
The full scope internal events (including internal flooding) PRA peer review was subject to a self-assessment and a full scope peer review conducted in March 2010 against ASME/ANS RA-Sa-2009 (Reference 9), Addenda A to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME and the American Nuclear Society, December 2008 (Reference 9), and any clarifications and qualifications provided in the NRC endorsement of the Standard contained in Revision 2 to RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, March 2009 (Reference 8).
A finding closure review was conducted on the internal events (including internal flooding) in October 2018. Closed findings were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, Close-out of Facts and Observations (F&Os) (Reference 10) as accepted by the NRC in the letter dated May 3, 2017 (ML17079A427)
(Reference 11). All Finding level F&Os were closed out.
The fire PRA (FPRA) was subject to a self-assessment and a full scope peer review conducted in October 2011 by the Pressurized Water Reactors Owners Group (PWROG) against all technical elements in Section 4 of ASME/ANS RA-Sa-2009 (Reference 9) and any clarifications and qualifications provided in RG 1.200, Revision 2 (Reference 8). After the full scope peer review, two focused-scope peer reviews (February 2018 and July 2018) and one Appendix X closure review (September 2018) were held. In the September 2018 closure review, all F&Os from the October 2011 full scope peer review, all F&Os from the February 2018 focused-scope peer review, and all F&Os from the July 2018 focused scope peer review were closed out.
In November-December 2019, a focused-scope peer review of the Farley internal events, internal flooding, fire, and seismic PRAs against applicable requirements of the ASME/ANS PRA standard (Reference 9) was conducted of the following PRA model upgrades:
x Reactor Coolant Pump Shutdown Seal Model (applicable to both internal events and internal flooding PRA models) x Main Control Room Abandonment (applicable to fire PRA model only) x FLEX modeling with FLEX HRA (applicable to all PRA models)
The FLEX HRA portion of the review was performed using the Integrated Human Event Analysis System (IDHEAS) Method based on Electric Power Research Institute (EPRI) FLEX Human Reliability Analysis (HRA) Report 3002013018 (Reference 12). As a result of this focused-scope peer review, a total of seven F&Os were generated, all of which were Suggestion type F&Os.
A Focused Scope Peer Review (FSPR) was performed in January 2023, against RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for
Enclosure to NL-25-0231 Basis for Proposed Changes E-6 Risk-Informed Activities," Revision 3 (Reference 13) to review some significant changes to the Farley PRA. These included revisions to the Fire PRA to incorporate the updated methods provided in NUREG-2230 and NUREG-2178. In the May 2023 Appendix X closure review, all F&Os from the January 2023 FSPR were closed out.
RCS Subcooling Margin is provided to determine SI termination and reinitiation and depressurization and cooldown progression. The RCS SMM measures saturation/superheat margin. The function of the RCS SMM is to calculate the subcooled margin which is the difference between the measured temperature of the reactor coolant and the saturation temperature. The saturation temperature is calculated from the minimum primary system pressure input. A maximum or representative temperature input is used for the measured value, which could come from a Resistance Temperature Detector (RTD) loop, or a representative core exit thermocouple (CET). Since the RCS SMM is not relied upon to support any PRA-modeled system to mitigate core damage or large early release, the RCS SMM has no contribution to the total Core Damage Frequency (CDF) and Large Early Release Frequency (LERF). Due to no contribution to CDF and LERF, the RCS SMM and any components associated with the RCS SMM are not High Safety Significant (HSS) or Low Safety Significant (LSS) with respect to Internal Events, Internal Flooding, Fire, or Seismic risk hazards.
3.4 Severe Accident Management Guidelines (SAMGs)
RCS Subcooling Margin is not used in the SAMGs as a primary indication and is not expected to be a reliable indication after significant core damage has occurred. The FNP SAMG technical support guideline for instrumentation states that subcooling margin monitor is not accurate after significant core damage. It does, however, list the RCS SMM as a secondary instrument that could be available for core exit thermocouples.
3.5 Emergency Plan RCS Subcooling Margin is indirectly used in emergency classification. The Critical Safety Function for core cooling (potential loss of the fuel clad barrier on the fission product barrier matrix) includes RCS subcooling and CET indications for orange entry.
3.6 Additional Plant-Specific Information Section 4.1 of the NRC SER of WCAP-15981-NP-A requires that amendment requests based on WCAP-15981-NP-A submit additional plant-specific information. The applicability of these additional requirements and their satisfaction is discussed below.
- 1. A general description of the PRA, including the scope of the analyses, PRA update history (including version peer reviewed, version(s) in which peer review comments were addressed, and version used for PAM application), and the licensee's PRA updating and quality assurance process.
- 2. A description of the most relevant peer reviews, a characterization of the peer review findings, a summary of the status of resolution of the peer review comments, and a listing of all unresolved facts and observations that potentially impact the application of TR WCAP-15981-NP.
Enclosure to NL-25-0231 Basis for Proposed Changes E-7
- 3. A conclusion regarding PRA quality assessment for the PAM TS application, and verification that the quality is acceptable for the application. This should include confirmation that the PRA reflects the as-built, as-operated design, and that any recent plant modifications and operational changes not reflected in the PRA do not impact the plant-specific PAM application; all peer review comments have been resolved or don't impact plant-specific PAM application; the PRA and HRA is sufficiently complete and applicable for evaluating the risk associated with the PAM application.
- 4. Listings of the important operator actions identified based on RAW and FV importance values for CDF and for LERF, along with these values.
- 5. Additions to the list of important operator actions based on review of results from the plant-specific external event assessments, or verification that the plant-specific risk assessments do not result in identification of additional risk-significant operator actions or variables/instruments.
- 6. A listing of variables/instruments related to the important operator actions. This should indicate how each variable/instrument considered in the methodology application was related to or mapped to a PRA model element or operator action.
- 7. Summary tables showing important indications for accident management, and the context in which they are important (e.g., DBA analysis, OBA, EOPs, SAMGs, PRA, EP (similar to Tables 7 and 8 in TR WCAP-15981-NP)).
- 8. A summary table describing variables/instruments added to or relocated from the technical specifications, and the specific bases for each change.
- 9. For any variables/instruments to be deleted from the TSs based on their lack of risk significance, the results of the focused evaluation of the adequacy of the PRA and HRA treatment (or lack of treatment) of operator actions associated with those variables/instruments.
- 10. For any variables/instruments to be deleted from the TSs based on their lack of risk significance, a discussion of how the reliability and availability of these instruments will be monitored and assessed (e.g., under the maintenance rule, other licensee program, or performance measurement strategy).
- 11. Prior to changing the treatment requirements for instrumentation that is used in emergency action level classification, licensees must consider the impacts of the changes on the effectiveness of their emergency plans.
Requirements 1, 2, 3, and 4 are related to the FNP PRA and are discussed in Section 3.3, Probabilistic Risk Assessment. The proposed change to the Subcooling Margin Monitor is not risk-informed or reliant on risk-insights.
Requirement 6 is related to the important operator actions related to the RCS Subcooling Margin and is discussed in Sections 3.1 and 3.2 of this enclosure.
Requirements 5, 7, 8, 9, 10, and 11 are not applicable to the proposed change discussed in this enclosure. Requirement 5 is not applicable because the proposed amendment does not add
Enclosure to NL-25-0231 Basis for Proposed Changes E-8 important operator actions and is not risk-based in nature. Requirements 7 and 8 are unnecessary since this amendment request contains a single PAM function change.
Requirements 9 and 10 are not applicable because this amendment request is not risk-based in nature. This amendment request is instead based upon the reclassification of the RCS Subcooling Margin to better identify its purpose in post accident monitoring at FNP. No variables/instruments are being deleted from the TSs based on their lack of risk significance.
Requirement 11 is not applicable as no treatment requirements for instrumentation used in emergency action level classification are being changed.
3.7 Removal of the RCS Subcooling Margin Monitor Function Contained in Technical Specification 3.3.3 The purpose of PAM instrumentation is to function in a post accident environment to provide indications necessary for operators to take manual actions to mitigate the consequences of an accident. PAM instrumentation may also include indications that have been determined to be risk significant. 10 CFR 50.36 (c)(2)(ii) Criterion 1 does not apply to the RCS SMM Function because it is not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. 10 CFR 50.36 (c)(2)(ii) Criterion 2 does not apply to the RCS SMM Function because it is not 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. Therefore, only Criteria 3 and 4 of 10 CFR 50.36 (c)(2)(ii) are applicable when evaluating instruments for retention in the PAM Technical Specification.
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.
Criterion 4: A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
The Standard Technical Specifications (STS) Bases for Westinghouse Plants B 3.3.3, "Post Accident Monitoring (PAM) Instrumentation" provides the following:
The instrument channels required to be OPERABLE by this LCO include two classes of parameters identified during unit specific implementation of Regulatory Guide 1.97 as Type A and Category I variables.
Type A variables are included in this LCO because they provide the primary information required for the control room operator to take specific manually controlled actions for which no automatic control is provided, and that are required for safety systems to accomplish their safety functions for DBAs.
Category I variables are the key variables deemed risk significant because they are needed to:
x Determine whether other systems important to safety are performing their intended functions,
Enclosure to NL-25-0231 Basis for Proposed Changes E-9 x
Provide information to the operators that will enable them to determine the likelihood of a gross breach of the barriers to radioactivity release, and x
Provide information regarding the release of radioactive materials to allow for early indication of the need to initiate action necessary to protect the public, and to estimate the magnitude of any impending threat.
Therefore, in addition to evaluating the RCS SMM Function against 10 CFR 50.36 (c)(2)(ii)
Criteria 3 and 4, the RCS SMM Function was evaluated based on the methodology in WCAP-15981-NP-A to determine the applicable Regulatory Guide 1.97 Type and Category.
The RCS subcooling indication provides information to indicate whether the core cooling safety function is being accomplished. This will not be changed. Type B variables are defined as those variables to be monitored that provide the control room operator with information to assess the process of accomplishing or maintaining critical safety functions. Therefore, for the purpose of determining the content of Technical Specification 3.3.3, "Post Accident Monitoring (PAM)
Instrumentation," this variable is considered a Type B variable. The RCS subcooling indication is a backup to the CETs and RCS pressure. The FNP response to Regulatory Guide 1.97 categorized the RCS SMM Function as a Category 2 variable. RG 1.97 Category I provides the most stringent PAM requirements and is intended for key variables. Category 2 provides less stringent requirements and generally applies to instrumentation designated for indicating system operating status. Therefore, for the purpose of determining the content of Technical Specification 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," the RCS SMM Function is considered a Category 2 variable.
Using the methodology of WCAP-15981-NP-A, it was determined that the RCS SMM does not fulfill the function of Regulatory Guide 1.97 for a Type A or Category I instrument. The evaluations performed in accordance with WCAP-15981-NP-A also support the conclusions that the RCS SMM does not meet 10 CFR 50.36 (c)(2)(ii) Criterion 3 (i.e., it was not found to be a Type A instrument) or Criterion 4 (i.e., it was not found to be significant to risk).
The inputs to the RCS SMM are RCS hot leg and cold leg temperatures from loop RTDs, CET temperature, RCS wide range pressure, and pressurizer pressure. Since these indications are independently displayed in the control room, the RCS SMM provides redundant calculation and display functions as a backup of these input indications. Therefore, the RCS SMM does not satisfy either Criterion 3 or 4 of 10 CFR 50.36 (c)(2)(ii) and should not be included in PAM Technical Specification 3.3.3.
The core subcooling monitor (RCS SMM) is currently listed as Regulatory Guide 1.97 Type A, Category 2 in UFSAR Table 7.5-1 (Sheet 1 of 16). Based on the methodology in WCAP-15981-NP-A, the RCS subcooling margin indication was determined to fulfill the post accident monitoring function of a Regulatory Guide 1.97 Type B variable for FNP Units 1 and 2. As such, the RCS SMM indication will be removed from TS Table 3.3.3-1, and will be retained in Farley UFSAR Table 7.5-1 (Sheet 3 of 16) as a Type B, Category 2 variable. Thereafter, changes to the RCS SMM function will be controlled in accordance with 10 CFR 50.59. The provisions of 10 CFR 50.59 establish adequate controls over requirements removed from the TS and assure future changes to these requirements will continue to be consistent with safe plant operation.
Enclosure to NL-25-0231 Basis for Proposed Changes E-10
4.0 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met.
GDC 13 includes a design requirement that instrumentation be provided to monitor variables and systems over their anticipated ranges for normal operation, anticipated operational occurrences, and accident conditions as appropriate to ensure adequate safety. There is no impact on the requirement of GDC 13 since the proposed amendment does not include any plant design changes. The proposed change does not eliminate or otherwise alter any existing instrumentation.
GDC 19 includes a requirement that a control room be provided from which actions can be taken to maintain the nuclear power unit in a safe condition under accident conditions, including loss-of-coolant accidents, and that equipment, including the necessary instrumentation, at appropriate locations outside the control room be provided with a design capability for prompt hot shutdown of the reactor. There is no impact on the requirements of GDC 19, since the proposed change does not include any design changes or plant modifications.
10 CFR 50.36 contains requirements applicable to the content of a plant's Technical Specifications. The proposed change utilizes the criteria of 10 CFR 50.36 (c)(2)(ii) to evaluate the content of the PAM Technical Specification. The proposed change includes the relocation of an instrument from the Technical Specifications that does not satisfy any of the criteria of 10 CFR 50.36 (c)(2)(ii). As such, the proposed change is consistent with the requirements of 10 CFR 50.36.
10 CFR 50.47 contains requirements for Emergency Plans. The methodology in WCAP-15981-NP-A utilizes the Emergency Plan to determine the instrumentation that should be included in the Technical Specifications. Therefore, the proposed change does not result in any changes to the Emergency Plan and does not result in a decrease in the effectiveness of the Emergency Plan.
10 CFR 50.49 specifies design and performance requirements for safety-related instrumentation exposed to adverse environments during accident conditions. The proposed change does not impact the requirements of 10 CFR 50.49. The proposed change does not introduce changes that affect the design or performance of the safety-related instrumentation subject to 10 CFR 50.49. There are no plant design changes or modifications associated with the proposed change.
In summary, the proposed change does not involve any design changes to the PAM instrumentation, or changes to the physical arrangement of PAM instrumentation. The proposed change results in altering the scope of the existing PAM Technical Specification consistent with the requirements of 10 CFR 50.36. Therefore, the proposed change maintains that the required PAM instruments remain capable of performing their post accident monitoring function. Thus, the proposed change does not adversely impact the design or performance characteristics of the PAM system or any other system.
Enclosure to NL-25-0231 Basis for Proposed Changes E-11 4.2 Precedent Letter M. B. Sellman (Entergy) to U.S. Nuclear Regulatory Commission, June 27, 1996, Technical Specification Change Request NPF-38-175 [ADAMS Accession No. ML20113C527].
4.3 No Significant Hazards Consideration Analysis Southern Nuclear Operating Company (SNC) requests a license amendment to the Joseph M.
Farley Nuclear Plant (FNP) Units 1 and 2 Renewed Facility Operating Licenses NPF-2 and NPF-8 respectively. The proposed amendment would revise Table 3.3.3-1 of the Technical Specifications (TS) to delete Function 10, RCS Subcooling Margin Monitor.
SNC has evaluated if a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
- 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No. The proposed change does not involve any changes to plant equipment, system design functions or a change in the methods governing normal plant operation. Therefore, the probability of a malfunction of a structure, system or component to perform its design function will not be significantly increased. The proposed change relocates specific instrumentation that does not fulfill the functional requirements of Regulatory Guide 1.97 Type A or Category I instrumentation, as evaluated in accordance with the methodology of WCAP-15981-NP-A. The proposed change also makes the Post Accident Monitoring (PAM) Instrumentation Technical Specification more consistent with the corresponding Westinghouse Plants Standard Technical Specifications (STS) PAM Technical Specification. Changes that are consistent with the STS have been evaluated and found not to adversely affect the safe operation of Westinghouse plants or initiate any accident previously evaluated.
Based on the conclusions of the plant-specific evaluation associated with the change and the evaluation performed in developing the STS, the proposed change does not result in operating conditions that will significantly increase the probability of initiating an analyzed event. The PAM instruments that are assumed to provide the information for manual operator actions for which no automatic control is provided are retained in the Technical Specifications. As a result, the consequences of any accident previously evaluated are not significantly increased.
The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained. The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Furthermore, the proposed change does not increase the types or amounts of radioactive effluent that may be released offsite or significantly increase individual or cumulative
Enclosure to NL-25-0231 Basis for Proposed Changes E-12 occupational/public radiation exposures. The proposed change is consistent with the safety analysis assumptions and resultant consequences. It is concluded that the change does not significantly increase the probability of occurrence of a malfunction of equipment important to safety.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No. The proposed change does not involve any changes to plant equipment, system design functions or the methods governing normal plant operation. No new accident initiators are introduced by these changes. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. The proposed change does not result in a change in the manner in which the PAM instruments provide plant information. There are no design changes associated with the license amendment. The proposed change does not change any existing accident scenarios, nor create any new or different accident scenarios. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the change does not impose any new or different operating requirements or eliminate any existing requirements.
The change does not alter assumptions made in the safety analysis. The proposed change is consistent with the safety analysis assumptions and current plant operating practice.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No. The change does not involve a significant reduction in a margin of safety. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined.
The safety analysis acceptance criteria are not impacted by the change. The proposed change does not include any plant design changes. The proposed change will not revise the indication provided by the affected instruments, and all operator actions based on these indications that are credited in the accident analyses will remain the same. As such, the proposed change will not result in plant operation in a configuration outside the design basis or assumptions of the design basis accident analyses.
Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.
Based on the above, SNC concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
Enclosure to NL-25-0231 Basis for Proposed Changes E-13 4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
5.0 ENVIRONMENTAL CONSIDERATION
A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
6.0 REFERENCES
- 1. Regulatory Guide 1.97, Revision 2
- 2. Supplement 1 to NUREG-0737, "TMI Action Items"
- 3. A-181866 Unit 1 RG 1.97 Compliance Review A-204866 Unit 2 RG 1.97 Compliance Review NRC SER for FNP RG 1.97 Compliance Report, Letter, Reeves to McDonald, 1/7/87, 2/12/87 (ML20211D523, ML20211D534, ML20211D508).
- 4. WCAP-15981-NP-A, "Post Accident Monitoring Instrumentation Re-Definition for Westinghouse NSSS Plants"
- 5. NUREG-1431, Revision 5.0, Standard Technical Specifications Westinghouse Plants, Volume 2, Bases.
- 6. Electric Power Research Institute (EPRI) 3002017583, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization.
- 7. Generic Letter 88-20, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), Supplement 4," USNRC, June 1991.
- 8. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, March 2009.
- 9. ASME/ANS RA-S-2009, Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, February 2009.
- 10. NEI Letter to NRC, "Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os)," February 21, 2017, Accession Number ML17086A431.
Enclosure to NL-25-0231 Basis for Proposed Changes E-14
- 11. NRC Letter to Mr. Greg Krueger (NEI), "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 7-12, and 12-13, Close Out of Facts and Observations (F&Os)," May 3, 2017, Accession Number ML17079A427.
- 12. Electric Power Research Institute (EPRI) Report 3002013018, "Human Reliability Analysis (HRA) for Diverse and Flexible Mitigation Strategies (FLEX) and Use of Portable Equipment: Examples and Guidance," November 30, 2018.
- 13. RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, December 2020.
to NL-25-0231 Joseph M. Farley Nuclear Plant - Units 1 & 2 License Amendment Request Proposing Removal of RCS Subcooling Margin Monitor from Technical Specifications List of Post Accident Monitoring Instrumentation Proposed Technical Specification Changes (Marked-up Pages)
PAM Instrumentation 3.3.3 Farley Units 1 and 2 3.3.3-3 Amendment No. 167 (Unit 1)
Amendment No. 159 (Unit 2)
Table 3.3.3-1 (page 1 of 1)
Post Accident Monitoring Instrumentation FUNCTION REQUIRED CHANNELS CONDITION REFERENCED FROM REQUIRED ACTION D.1
- 1. RCS Hot Leg Temperature (Wide Range) 2 E
- 2. RCS Cold Leg Temperature (Wide Range) 2 E
- 3. RCS Pressure (Wide Range) 2 E
- 4. Steam Generator (SG) Water Level (Wide or Narrow Range) 2/SG E
- 5. Refueling Water Storage Tank Level 2
E
- 6. Containment Pressure (Narrow Range) 2 E
- 7. Pressurizer Water Level 2
E
- 8. Steam Line Pressure 2/SG E
- 9. Auxiliary Feedwater Flow Rate 2
E
- 10. RCS Subcooling Margin Monitor 2
E
- 11. Containment Water Level (Wide Range) 2 E
- 12. Core Exit Temperature - Quadrant 1 2(a)
E
- 13. Core Exit Temperature - Quadrant 2 2(a)
E
- 14. Core Exit Temperature - Quadrant 3 2(a)
E
- 15. Core Exit Temperature - Quadrant 4 2(a)
E
- 16. Reactor Vessel Level Indicating System 2
F
- 17. Condensate Storage Tank Level 2
E
- 18. Deleted
- 19. Containment Area Radiation (High Range) 2 F
(a) A channel consists of two core exit thermocouples.
Deleted NL-25-0231, Attachment 1 Page 1 of 1 to NL-25-0231 Joseph M. Farley Nuclear Plant - Units 1 & 2 License Amendment Request Proposing Removal of RCS Subcooling Margin Monitor from Technical Specifications List of Post Accident Monitoring Instrumentation Revised Technical Specification Pages (Clean Typed)
PAM Instrumentation 3.3.3 Farley Units 1 and 2 3.3.3-3 Amendment No. (Unit 1)
Amendment No. (Unit 2)
Table 3.3.3-1 (page 1 of 1)
Post Accident Monitoring Instrumentation FUNCTION REQUIRED CHANNELS CONDITION REFERENCED FROM REQUIRED ACTION D.1
- 1. RCS Hot Leg Temperature (Wide Range) 2 E
- 2. RCS Cold Leg Temperature (Wide Range) 2 E
- 3. RCS Pressure (Wide Range) 2 E
- 4. Steam Generator (SG) Water Level (Wide or Narrow Range) 2/SG E
- 5. Refueling Water Storage Tank Level 2
E
- 6. Containment Pressure (Narrow Range) 2 E
- 7. Pressurizer Water Level 2
E
- 8. Steam Line Pressure 2/SG E
- 9. Auxiliary Feedwater Flow Rate 2
E
- 10. Deleted
- 11. Containment Water Level (Wide Range) 2 E
- 12. Core Exit Temperature - Quadrant 1 2(a)
E
- 13. Core Exit Temperature - Quadrant 2 2(a)
E
- 14. Core Exit Temperature - Quadrant 3 2(a)
E
- 15. Core Exit Temperature - Quadrant 4 2(a)
E
- 16. Reactor Vessel Level Indicating System 2
F
- 17. Condensate Storage Tank Level 2
E
- 18. Deleted
- 19. Containment Area Radiation (High Range) 2 F
(a) A channel consists of two core exit thermocouples.
NL-25-0231, Attachment 2 Page 1 of 1 to NL-25-0231 Joseph M. Farley Nuclear Plant - Units 1 & 2 License Amendment Request Proposing Removal of RCS Subcooling Margin Monitor from Technical Specifications List of Post Accident Monitoring Instrumentation Proposed Technical Specification Bases Pages (Marked-up) - For Information Only
PAM Instrumentation B 3.3.3 Farley Units 1 and 2 B 3.3.3-9 Revision 0 BASES LCO 10.
RCS Subcooling Margin Monitor (continued)
RCS subcooling is a Category II, Type A variable provided to determine safety injection termination and reinitiation and depressurization and cooldown progression. The subcooled margin monitor (SMM) measures saturation/superheat margin.
The function of the SMM is to calculate the subcooled margin which is the difference between the measured temperature of the reactor coolant and the saturation temperature. The saturation temperature is calculated from the minimum primary system pressure input. A maximum or representative temperature input is used for the measured value, which could come from an RTD loop, or a representative core exit thermocouple.
11.
Containment Sump Water Level (Wide Range)
Containment Sump Water Level is a Category I, Type A variable provided for verification and long term surveillance of RCS integrity. This information provides a diverse means for checking RWST level.
Containment Sump Water Level is used to determine:
containment sump level accident diagnosis; and when to begin the recirculation procedure.
12, 13, 14, 15.
Core Exit Temperature Core Exit Temperature is provided for verification and long term surveillance of core cooling.
Adequate monitoring of core cooling is ensured with two valid Core Exit Temperature channels per quadrant with two core exit thermocouples (CETs) per required channel. The CET pair are oriented radially to permit evaluation of core radial decay power distribution. Core Exit Temperature is used to determine whether to terminate SI, if still in progress, or to reinitiate SI if it has been stopped. Core Exit Temperature is also used for unit stabilization and cooldown control.
(continued)
Deleted NL-25-0231, Attachment 3 Page 1 of 2
NL-25-0231, Attachment 3 Page 2 of 2 to NL-25-0231 Joseph M. Farley Nuclear Plant - Units 1 & 2 License Amendment Request Proposing Removal of RCS Subcooling Margin Monitor from Technical Specifications List of Post Accident Monitoring Instrumentation Proposed Updated Final Safety Analysis Report Pages (Marked-up) - For Information Only
FNP-FSAR-7 7.5-4 REV 21 5/08 A maximum TH and a maximum T are selected and are used to calculate separate setpoint signals for the heater controllers. The minimum of the two heater controller setpoint signals is selected and sent to each of the heater controllers. The TH and T heater controller setpoint signals are reduced at a constant rate, when their respective TH and T values increase above a predetermined value. The TH and T heater controller setpoints will decrease until they equal zero at a second predetermined setpoint.
7.5.4.2 Subcooling Margin Monitor The subcooling margin monitor (SMM) provides continuous, redundant indication of the margin to saturated conditions in the reactor coolant system (RCS). The SMM inputs are RCS hot leg and cold leg temperatures from loop RTDs, core exit thermocouple temperature, RCS wide range pressure, and pressurizer pressure. The margin to saturation, displayed in degrees F, is the difference between the measured RCS temperature and the saturation temperature. The highest RCS loop temperature and the highest core exit thermocouple temperature, excluding upper head thermocouples, are used to calculate margins to saturation. The lowest pressure value is used to calculate the saturation temperature. The control board SMM display has a switch to select margin to saturation indication based on RCS loop RTD temperature or core exit thermocouple temperature.
7.5.4.3 Core Exit Temperature Core exit temperature is continuously indicated on redundant control board displays. The chromel-alumel thermocouples in the vessel measure temperature at the flow exit of selected fuel assemblies and locations within the reactor vessel head plenum.
The redundant displays each normally indicate the temperature of the hottest thermocouple for that channel. The operator can interrogate the display to indicate the temperature of any individual thermocouple or the highest temperature in each core quadrant.
7.5.5 NUCLEAR INSTRUMENTATION In addition to the Westinghouse nuclear instrumentation system that is described in section 7.2 and whose indications are listed in table 7.5-3, an independent channel of Gamma-Metrics nuclear instrumentation is provided to satisfy alternate shutdown requirements.
The Gamma-Metrics channel provides neutron flux indication at the hot shutdown panel and the control room via isolated outputs. A fission chamber detector measures neutron flux from shutdown to full power. Detector sensitivity is 10-2 to 1010 nv. The following displays are provided on the main control board and the hot shutdown panel:
Amendments XXX and YYY to the Facility Operating Licenses for Units 1 and 2, respectively,removed the Subcooling Margin Monitor and related administrative controls from the Technical Specifications NL-25-0231, Attachment 4 Page 1 of 3
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