ML23318A067
| ML23318A067 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/31/2023 |
| From: | Coleman J Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML23318A074 | List: |
| References | |
| NL-23-0806 | |
| Download: ML23318A067 (19) | |
Text
Enclosure 3 contains security-related information and should be withheld under 1 O CFR 2.390.
Upon removal of Enclosure 3, this correspondence is suitable for public disclosure.
~ Southern Nuclear October 31, 2023 Docket Nos.: 50-348 50-364 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Jamie M. Coleman Regulatory Affairs Director Joseph M. Farley Nuclear Plant - Units 1 & 2 3535 Colonnade Parkway Birmingham, AL 35243 205 992 6611 tel 205 992 7795 fax jamiemco@southernco.com NL-23-0806 Revision 31 to the Updated Final Safety Analysis Report, Updated NFPA 805 Fire Protection Program Design Basis Document, Technical Specification Bases Changes, Technical Requirements Manual Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report Ladies and Gentlemen:
In accordance with 10 CFR 50.4(b) and 50.71 (e), Southern Nuclear Operating Company (SNC) hereby submits Revision 31 to the Joseph M. Farley Nuclear Plant (FNP) Updated Final Safety Analysis Report (UFSAR). The revised FNP UFSAR, indicated as Revision 30, reflects changes through October 15, 2023.
The FNP Technical Specifications, Section 5.5.14, "Technical Specifications (TS) Bases Control Program," provides for changes to the Bases without prior Nuclear Regulatory Commission (NRC) approval. In addition, TS Section 5.5.14 requires that Bases changes made without prior NRC approval be provided to the NRC on a frequency consistent with 10 CFR 50.71(e}. Pursuant to TS 5.5.14, SNC hereby submits a complete copy of the FNP TS Bases. The revised FNP TS Bases, indicated as Revision 116, reflects changes to the TS Bases through October 15, 2023.
The revised FNP TRM, indicated as Version 61.0, reflects changes to the TRM through October 15, 2023. The updated National Fire Protection Association (NFPA) 805 Fire Protection Program Design Basis Document, Version 7.0, also reflects changes through October 15, 2023.
In accordance with Regulatory Issue Summary (RIS) 2001 -05, "Guidance on Submitting Documents to the NRC by Electronic Information Exchange or on CD-ROM," all of the current pages of the FNP UFSAR, the FNP UFSAR reference drawings, the TS Bases, the Technical Requirements Manual (TRM), and the NFPA 805 Fire Protection Program Design Basis Document are hereby submitted on the Electronic Information Exchange in portable document format (PDF).
In accordance with 1 O CFR 50.59(d)(2), SNC hereby submits the 1 O CFR 50.59 Summary Report containing a brief description of any changes, tests, or experiments, including a summary of the safety evaluation of each.
U.S. Nuclear Regulatory Commission NL-23-0806 Page 2 In accordance with NEI 99-04, "Guidelines for Managing NRC Commitment Changes,"
Revision 0, SNC hereby submits a summary of commitment changes for the applicable reporting period (October 15, 2021 to October 15, 2023).
SNC conducted a review of FNP plant changes for 10 CFR 54.37(b) applicability and identified no components that were determined to meet the criteria for newly identified components as clarified by RIS 2007-16, Revision 1, "Implementation of the Requirements of 10 CFR 54.37(b) for Holders of Renewed Licenses." provides a table of contents with associated file names for the Electronic Information Exchange (EIE) (Enclosure 2, public version with security-related information redacted; Enclosure 3, non-public version). contains security-related information within the Farley UFSAR. SNC requests that Enclosure 3 be withheld from public disclosure in its entirety in accordance with 10 CFR 2.390(d)(1). Enclosure 4 provides the 10 CFR 50.59 Summary Report. provides the Revised NRC Commitments Report This letter contains no NRC commitments. If you have any questions, please contact Ryan Joyce at (205) 992-6468.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on this 31 st day of October 2023.
Respectfully submitted,
~~
Jamie Coleman Regulatory Affairs Director JMC/rdh/cg
Enclosures:
- 1. Electronic Information Exchange Table of Contents
- 2. Electronic Information Exchange - Public documents
- 3. Electronic Information Exchange - Nonpublic documents (Withhold from public disclosure in accordance with 1 O CFR 2.390( d)( 1))
- 4. 10 CFR 50.59 Summary Report
- 5. Revised NRC Commitments Report cc: Regional Administrator, Region II (w/o enclosures)
Senior NRR Project Manager - Farley (w/o enclosures)
Senior Resident Inspector-Farley (w/o enclosures)
RType: CFA04.054
NL-23-0806 Joseph M. Farley Nuclear Plant - Units 1 & 2 Revision 31 to the Updated Final Safety Analysis Report, Updated NFPA 805 Fire Protection Program Design Basis Document, Technical Specification Bases Changes, Technical Requirements Manual Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report Electronic Information Exchange Table of Contents
I to NL-23-0806 Electronic Information Exchange Table of Contents FILENAME SEQ CONTENT FOR PUBLIC ENCLOSURE 2*
EXTENSION 001 EPL-TOC_P
.pdf Effective Page List and Table of Contents 002 Chapter 1_P
.pdf 003 Chapter 2_P
.pdf 004 Chapter 3_P
.pdf 005 Chapter 4_P
.pdf 006 Chapter 5_P
.pdf 007 Chapter 6_P
.pdf 008 Chapter 7_P
.pdf 009 Chapter 8_P
.pdf 010 Chapter 9_P
.pdf 011 Chapter 10_P
.pdf 012 Chapter 11_P
.pdf 013 Chapter 12_P
.pdf 014 Chapter 13_P
.pdf 015 Chapter 14_P
.pdf 016 Chapter 15_P
.pdf 017 Chapter 16_P
.pdf 018 Chapter 17 _P
.pdf 019 Chapter 18_P
.pdf 027*
TECHNICAL SPECIFICATIONS BASES_P
.pdf 028*
TECHNICAL REQUIREMENTS MANUAL_P
.pdf 029*
NFPA 805 FIRE PROTECTION PROGRAM_P
- Files 020 through 026 are completely non-public and thus not provided on the Enclosure 2 public electronic information exchange (EIE) submittal. Files 027 through 029 are completely public and thus not provided on the Enclosure 3 non-public EIE submittal.
E1-1 to NL-23-0806 Electronic Information Exchange Table of Contents I
FILENAME SEQ CONTENT FOR NON-PUBLIC ENCLOSURE 3*
EXTENSION 001 EPL-TOC_NP
.pdf Effective Page List and Table of Contents 002 Chapter 1_NP
.pdf 003 Chapter 2_NP
.pdf 004 Chapter 3_NP
.pdf 005 Chapter 4_NP
.pdf 006 Chapter 5_NP
.pdf 007 Chapter 6_NP
.pdf 008 Chapter 7 _NP
.pdf 009 Chapter 8_NP
.pdf 010 Chapter 9_NP
.pdf 011 Chapter 10_NP
.pdf 012 Chapter 11_NP
.pdf 013 Chapter 12_NP
.pdf 014 Chapter 13_NP
.pdf 015 Chapter 14_NP
.pdf 016 Chapter 15_NP
.pdf 017 Chapter 16_NP
.pdf 018 Chapter 17 _NP
.pdf 019 Chapter 18_NP
.pdf 020*
FARLEY FSAR REF DWGS_NP
.pdf A 177048 sh 1 thru A 177048 sh 325 021*
FARLEY FSAR REF DWGS_NP
.pdf A 177048 sh 326 thru A 177048 sh 568 022*
FARLEY FSAR REF DWGS_NP
.pdf A207048 sh 1 thru A207048 sh 300 E1-2
I to NL-23-0806 Electronic Information Exchange Table of Contents FILENAME SEQ CONTENT FOR NON-PUBLIC ENCLOSURE 3*
EXTENSION 023*
FARLEY FSAR REF OWGS_NP
.pdf A207048 sh 301 thru A207048 sh 568 024*
FARLEY FSAR REF OWGS_NP
.pdf A508650 sh 1 thru 0175012 sh 1 025*
FARLEY FSAR REF OWGS_NP
.pdf 0175014 sh 1 thru 0177944 sh 1 026*
FARLEY FSAR REF OWGS_NP
.pdf 0181620 sh 1 thru U611138
- Files 020 through 026 are completely non-public and thus not provided on the Enclosure 2 public optical disc. Files 027 through 029 are completely public and thus not provided on the Enclosure 3 non-public optical disc.
E1-3
NL-23-0806 Joseph M. Farley Nuclear Plant - Units 1 & 2 Revision 31 to the Updated Final Safety Analysis Report, Updated NFPA 805 Fire Protection Program Design Basis Document, Technical Specification Bases Changes, Technical Requirements Manual Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report Electronic Information Exchange - Public
NL-23-0806 Joseph M. Farley Nuclear Plant - Units 1 & 2 Revision 31 to the Updated Final Safety Analysis Report, Updated NFPA 805 Fire Protection Program Design Basis Document, Technical Specification Bases Changes, Technical Requirements Manual Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report Electronic Information Exchange - Nonpublic (Withhold from public disclosure in accordance with 10 CFR 2.390(d)(1))
NL-23-0806 Joseph M. Farley Nuclear Plant - Units 1 & 2 Revision 31 to the Updated Final Safety Analysis Report, Updated NFPA 805 Fire Protection Program Design Basis Document, Technical Specification Bases Changes, Technical Requirements Manual Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report 10 CFR 50.59 Summary Report to NL-23-0806 10 CFR 50.59 Summary Report 10 CFR 50.59 Summary Report Activity: LDCR22-021
Title:
Technical Requirements Manual (TRM) Burden Reduction 10 CFR 50.59 Evaluation Summary:
This activity revises TR 13.1.2, "Boration Flow Path - Shutdown," TR 13.1.4, "Boration Pump -
Shutdown," and TR 13.1.6, "Borated Water Source - Shutdown," by simplifying the required Actions for these conditions by deleting the unnecessary action to suspend CORE ALTERATIONS and providing a single, more specific immediate action to "Suspend operations involving positive reactivity changes." A conforming change is also made to the TR Bases for TR 13.1.2 through TR 13.1.7. The changes associated with this activity are administrative changes to UFSAR procedural-like requirements for nonfunctional boration system equipment.
This activity revises TR 13.1. 7, "Borated Water Sources - Operating," by removing the requirement stating that the refueling water storage tank (RWST) is a required borated water source whenever in MODES 1, 2, 3, and 4. This revision involves changes to the TR requirement and deletion of Condition D, Surveillance Note 1, and TR Surveillance (TRS) 13.1.7.1 (and renumbering the remaining surveillances). Surveillance Note 2 is revised to clarify that the remaining surveillances for the boric acid storage tank are only required if this tank is a required borated water source, per TR 13.1.3. This activity involves an administrative change to UFSAR procedural-like requirements (TRM) such that Required Actions for the TRM are removed and reliance for the RWST as a borated water source is placed on RWST Technical Specification 3.5.4. A conforming change is made to the TR Bases for TR 13.1.2 through 13.1. 7.
This activity removes the requirements of TR 13.1.10, "Rod Drop Time," for the performance of a rod drop time test following maintenance on or modification to the control rod drive system that could affect the drop time of specific rods and adds text to TSB SR 3.1.4.3 to address the requirement deleted from the TRM. This change clarifies the licensing basis by deleting duplicate information, thereby improving reader understanding of the licensing basis without changing the meaning or substance of the information presented. The change to TR 13.1.10 and TSB SR 3.1.4.3 retains the basis for the current TRM requirement by identifying maintenance and modification of the control rod drive system as an activity that could affect the drop time of those specific rods. Conforming changes are made to the Bases for TR 13.1.1 O and UFSAR Subsection 4.2.3.4.2.
This activity removes TR 13.3.6, "Seismic Monitoring Instrumentation," Condition B, which provides the Actions to be performed following a seismic event involving the actuation of one or more seismic monitoring instruments. This activity includes a change to UFSAR procedural-like requirements (TRM) such that Required Actions to evaluate data from the seismic instruments to determine the magnitude of the vibratory ground motion and restore the instrumentation to service following a seismic event are eliminated, and only the more general actions remain to be followed whenever seismic monitoring instrumentation is determined to be nonfunctional.
E4-1 to NL-23-0806 10 CFR 50.59 Summary Report Removal of Condition B and the associated Required Actions does not adversely affect the capability of this instrumentation to provide its monitoring functions as described in the UFSAR.
This activity removes TRM requirements of TR 13.4.1, "Chemistry" requirements. The changes associated with the deletion of this TRM specification are administrative changes to UFSAR procedural-like requirements (TRM) to be taken which are the same as those actions required by the Corporate PWR Primary Water Chemistry Program and the associated implementing procedures that implement the EPRI PWR Primary Water Chemistry Guidelines, which provides the basis for the same chemistry limits in the TRM. The TR Bases is revised to show that TR 13.4.1, Chemistry, is not used. Conforming changes are made to UFSAR Subsections 5.5.1.1.1 and 5.5.12.3.
This activity clarifies and simplifies the administrative requirements in TR 13.4.5, "RCS Pressure Isolation Valve (PIV) Leakage," by deleting the requirements from the TRM that duplicate the requirements of TS SR 3.4.14.1, and retaining only the table that identifies the valves that are subject to TS 3.4.14, including TS SR 3.4.14.1. A conforming change is made to the TR Bases for TR 13.4.5 and a statement is added to TSB 3.4.14 to address post-work testing.
This activity removes the TRM requirements of TR 13.5.1, "Emergency Core Cooling System (ECCS)," requirement "a" for "Unrestricted containment sump suctions," and surveillance TRS 13.5.1.1, which provides high-level procedure-like requirements that address containment cleanliness visual inspection requirements as described in the UFSAR. TR 13.5.1.a and TRS 13.5.1.1 are to be identified as "Not used." Because containment cleanliness visual inspections are also required by TS Surveillance Requirement (SR) 3.6.10.1 and plant operating procedure requirements, deletion of this requirement from the TRM does not involve an adverse effect on the performance or method of control of a design function as described in the Updated FSAR.
This activity clarifies and simplifies the administrative requirements in TR 13.6.2, "Containment Isolation Valves (CIVs)," by deleting the requirements from the TRM that duplicate the requirements of TS SR 3.6.3.4, and retaining only the table identifies the valves that are subject to these TS requirements. A note is added stating that "Table 13.6.2-1 contains a listing of Containment Isolation Valves subject to Technical Specification SR 3.6.3.4" and a statement is added to the Bases for TS 3.6.3, "Containment Isolation Valves," referring to the TRM for the list of containment isolation valves. A conforming change is made to the TR Bases for TR 13.6.2.
This activity removes the TRM requirements of TR 13. 7.2, "Snubbers," and revises the TR 13.7.2 Bases to state "Not used." The changes associated with the TRM Specification are administrative changes to UFSAR procedural-like requirements, which are duplicative of other procedural requirements which implement the TS definition of Operability and TS 3.0.8 for supported systems.
The activities were not identified to more than minimally increase the frequency of occurrence or consequences of an accident previously evaluated in the Updated FSAR, more than minimally increase the likelihood of occurrence or consequences of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the Updated FSAR, create the possibility for an accident of a different type or for a malfunction of an SSC important to safety E4-2 to NL-23-0806 10 CFR 50.59 Summary Report with a different result than any previously evaluated in the Updated FSAR, have any impact on the integrity of the fuel cladding, reactor coolant pressure boundary, or containment, or result in a departure from a method of evaluation described in the Updated FSAR used in establishing the design bases or in the safety analyses.
Activity: DCP SNC1358093
Title:
Unit 2 Temporary Loop System (TLS) 10 CFR 50.59 Evaluation Summary:
The proposed activity upgrades the Westinghouse 7300-series nuclear steam supply system (NSSS) control systems by incorporating these into the existing Ovation-based TLS. Various control system changes reflecting improvements and lessons learned from operating experience with the existing control systems are also incorporated into the new design. Therefore, the proposed activity consists of both a digital upgrade and functional changes to these temporary control systems.
The control systems within the scope of the proposed activity include the following major control systems:
Pressurizer pressure control, Pressurizer level control Chemical and volume control (Boration/reactor makeup control & VCT level control)
RHR Heat Exchanger Bypass Flow control Failures in these control systems have the potential to initiate the following Condition II event previously evaluated in the Updated FSAR.
Uncontrolled boron dilution (15.2.4)
Since an occurrence of any of the accidents identified above as a result of the proposed activity would be due to a failure in the modified systems, the modified systems must exhibit a low likelihood of failure so that there is no more than a minimal increase in the frequency of occurrence of these accidents. Likewise, the modified systems must exhibit a low likelihood of failure so that there is no more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety.
A Qualitative Assessment (performed in accordance with Regulatory Issue Summary (RIS) 2002-22 and RIS 2002-22, Supplement 1) of factors involving the design attributes of the digital upgrade, the quality of the design process used in developing the upgrade, and operating experience with similar upgrades concluded that the digital upgrade will exhibit a sufficiently low likelihood of failure.
The functional changes or enhancements to the control systems were reviewed on a system-by-system basis in DCP SNC1358093; the discussion of differences there includes the following:
E4-3 to NL-23-0806 10 CFR 50.59 Summary Report a description of the existing 7300-series configuration of each control system, including the operator interface at the main control board identification of the changes associated with the Ovation-based TLS, including changes to the operator interface a review of the impact of the changes on the control system, including the impact on the operator interface and on the control system failure modes and effects On the basis of the Qualitative Assessment of the digital upgrade and the review of the functional changes, it was concluded that the proposed activity does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the Updated FSAR or in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the Updated FSAR.
Although failures in the major control systems can cause the previously evaluated accident identified above, failure modes and effects analyses performed in support of the proposed activity concluded that the existing failure effects remain bounding. Therefore, it was concluded that existing radiological dose consequences remain bounding, such that the proposed activity does not result in more than a minimal increase in the consequences of an accident previously evaluated in the Updated FSAR or in the consequences of a malfunction of an SSC important to safety previously evaluated in the Updated FSAR.
As described above, existing failures and failure modes of the major control systems could result in accidents and malfunctions previously described in the FSAR. It was recognized that the incorporation of multiple control systems into a network utilizing common resources introduced the potential to create an accident or malfunction not previously considered. The results of the Qualitative Assessment indicated that the likelihood of a software common cause failure was sufficiently low - that is, comparable to the likelihood of failures that are not considered in the Updated FSAR. On that basis, it was concluded that the proposed activity does not create a possibility for an accident of a different type or a malfunction with a different result than any previously evaluated in the Updated FSAR.
Since failures in the major control systems can cause the previously evaluated accidents identified above, and since the failure modes and effects analyses performed in support of the proposed activity determined that the existing failure effects remain bounding, the existing safety analyses, which demonstrate that design basis limits for a fission product barrier are not exceeded, remain bounding for failures in the modified control systems. Therefore, it was concluded that the proposed activity does not result in a design basis limit for a fission product barrier as described in the Updated FSAR being exceeded or altered.
Finally, no Updated FSAR-described method of evaluation is affected by the proposed activity.
Activity: DCP SNC1083455
Title:
U2 7300 and SGFP Control System Replacement 10 CFR 50.59 Evaluation Summary:
E4-4 to NL-23-0806 10 CFR 50.59 Summary Report The proposed activity upgrades the Westinghouse 7300-series nuclear steam supply system (NSSS) control systems, the steam generator water level (SGWL) control system, and the steam generator feedwater pump (SGFP) control system by incorporating these into the existing Ovation-based distributed control system (DCS). The existing Ovation-based DCS will also be modified to accommodate these control systems. Various control system changes reflecting improvements and lessons learned from operating experience with the existing control systems are also incorporated into the new design. Therefore, the proposed activity consists of both a digital upgrade and functional changes to the affected control systems.
The control systems within the scope of the proposed activity include the following major control systems:
Distributed control system Reactor coolant (reactor (Tavg) control, pressurizer pressure control, pressurizer level control)
Chemical and volume control (reactor makeup control)
Main steam (steam dump control)
Main feedwater (steam generator level control, turbine-driven feedwater pump speed control)
Main turbine (main turbine control)
Failures in these control systems have the potential to initiate the following Condition II events previously evaluated in the Updated FSAR.
Uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition (15.2.1)
Uncontrolled rod cluster control assembly bank withdrawal at power (15.2.2)
Uncontrolled boron dilution (15.2.4)
Turbine trip (15.2.7)
Loss of normal feedwater (15.2.8)
Excessive heat removal due to feedwater system malfunctions (15.2.10)
Excessive load increase (15.2.11)
Accidental depressurization of the RCS (15.2.12)
Accidental depressurization of the main steam system (15.2.13)
Since an occurrence of any of the accidents identified above as a result of the proposed activity would be due to a failure in the modified systems, the modified systems must exhibit a low likelihood of failure so that there is no more than a minimal increase in the frequency of occurrence of these accidents. Likewise, the modified systems must exhibit a low likelihood of failure so that there is no more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety.
A qualitative assessment (performed in accordance with Regulatory Issue Summary (RIS) 2002-22 and RIS 2002-22 Supplement 1) of factor involving the design attributes of the digital upgrade, the quality of the design process used in developing the upgrade, and operating E4-5 to NL-23-0806 10 CFR 50.59 Summary Report experience with similar upgrades concluded that the digital upgrade will exhibit a sufficiently low likelihood of failure.
The functional changes or enhancements to the control systems were reviewed on a system-by-system basis in DCP SNC1083455; the discussion of differences there includes the following:
A description of the existing 7300-series configuration of each control system, including the operator interface at the main control board Identification of the changes associated with the Ovation upgrade to the control system, including changes to the operator interface If not clear from the Ovation upgrade summary, a description of the impact of the changes on the control system, including the impact on the operator interface The impact on the control system failure modes and effects is documented in the vendor-furnished FMEA (WNA-AR-00959) and an additional FMEA (SNC1083455-J001) produced to cover the modification scope not detailed in the vendor-furnished FMEA. The acceptability of these FMEA results demonstrate that the proposed activity does not represent a modification, addition, or removal of a SSC such that a design function as described in the Updated FSAR is adversely affected.
On the basis of the qualitative assessment of the digital upgrade and the review of the functional changes, it was concluded that the proposed activity does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the Updated FSAR or in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the Updated FSAR.
Although failures in the major control systems can cause the previously evaluated accidents identified above, failure mode and effects analyses performed in support of the proposed activity concluded that the existing failure effects remain bounding. Therefore, it was concluded that existing radiological dose consequences remain bounding, such that the proposed activity does not result in more than a minimal increase in the consequences of an accident previously evaluated in the Updated FSAR or in the consequences of a malfunction of an SSC important to safety previously evaluated in the Updated FSAR.
As described above, existing failures and failure modes of the major control systems could result in accidents and malfunctions previously described in the FSAR. It was recognized that the incorporation of multiple control systems into a network utilizing common resources introduced the potential to create an accident or malfunction not previously considered. The results of the qualitative assessment indicated the likelihood of a common cause failure was sufficiently low - that is, comparable to the likelihood of failures that are not considered in the Updated FSAR. On that basis, it was concluded that the proposed activity does not create a possibility for an accident of a different type or a malfunction with a different result than any previously evaluated in the Updated FSAR.
Since failures in the major control systems can cause the previously evaluated accidents identified above, and since the failure modes and effects analyses performed in support of the proposed activity determined that the existing failure effects remain bounding, the existing E4-6 to NL-23-0806 10 CFR 50.59 Summary Report safety analyses, which demonstrate that design basis limits for a fission product barrier are not exceeded, remain bounding for failures in the modified control systems. Therefore, it was concluded that the proposed activity does not result in a design basis limit for a fission product barrier as described in the Updated FSAR being exceeded or altered.
Finally, no Updated FSAR-described method of evaluation is affected by the proposed activity.
E4-7
NL-23-0806 Joseph M. Farley Nuclear Plant - Units 1 & 2 Revision 31 to the Updated Final Safety Analysis Report, Updated NFPA 805 Fire Protection Program Design Basis Document, Technical Specification Bases Changes, Technical Requirements Manual Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report Revised NRC Commitments Report to NL-23-0806 Revised NRC Commitments Report Original Commitment: SNC500062: Extension of Operating License Duration The QA manager (formerly MSAER) will ensure that periodic audits of the corporate plant ALARA programs are conducted to ensure regulatory compliance. The QA Supervisor (formerly Supervisor of Safety Audit and Engineering Review (SSAER)), reporting to the QA Manager (formerly MSAER), shall audit the plant ALARA program at least once per 24 months annually.
These audits should ensure that the ALARA program continues to meet the intent described in LC#00728, (re: 40 years extension from full power license) plant life extension submittal.
Revised Commitment:
NOS Management will ensure that periodic audits of the ALARA Programs are conducted to ensure regulatory compliance. ALARA Program Reviews are conducted annually per NMP-AD-035 which directs the completion of the Annual ALARA Report, and NOS audits of the ALARA Program will be conducted during the Radiation Protection audit in accordance with the frequency documented in NOS-104.
Justification for Changes:
The requirements of 10 CFR 20.1101 (c) continue to be met. While the frequency of audits are being revised from at least once every 24-months to at least once every 36-months [per application of the NRC approval of Exelon's submittal as described in a Safety Evaluation dated November 5, 2020 (ML20287A130) via 10 CFR 50.54(a)(3)(ii)], the intent of the original commitment is still being maintained.
Additionally, the commitment is consistent with NRC's response to Health Physics Questions and Answers question 118.