NL-23-0008, Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specification 3.4.10, Pressurizer Safety Valves to Decrease Low Side Setpoint Tolerance Limiting.
| ML23017A205 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 01/17/2023 |
| From: | Gayheart C Southern Nuclear Operating Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| NL-23-0008 | |
| Download: ML23017A205 (1) | |
Text
Cheryl A. Gayheart Regulatory Affairs Director 3535 Colonnade Parkway Birmingham, AL 35243 205 992 5316 cagayhea@southernco.com January 17, 2023 Docket Nos. 50-348 NL-23-0008 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Units 1 and 2 Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specification 3.4.10, Pressurizer Safety Valves to Decrease Low Side Setpoint Tolerance Limiting Condition for Operation (LCO) Value By letter dated June 30, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22181B145), Southern Nuclear Operating Company (SNC) submitted a license amendment request (LAR) for the Joseph M. Farley Nuclear Plant, Units 1 and 2 (FNP). The LAR proposes to revise FNP Technical Specification (TS) 3.4.10, Pressurizer Safety Valves to decrease the low side setpoint tolerance value from
- 1% to - 2.5%.
By email dated November 30, 2022 (ML22334A148), the U.S. Nuclear Regulatory Commission (NRC) notified SNC that additional information is needed for the staff to perform their review.
The enclosure to this letter provides the SNC response to the NRC Request for Additional Information (RAI).
The conclusions of the No Significant Hazards Consideration and Environmental Consideration contained in the original application have been reviewed and are unaffected by this response.
In accordance with 10 CFR 50.91, a copy of this application, including attachments, is being provided to the designated Alabama Official.
This letter contains no regulatory commitments. If you have any questions, please contact Ryan Joyce at 205.992.6468.
U. S. Nuclear Regulatory Commission NL-23-0008 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 17th day of January 2023.
Respectfully submitted, C. A. Gayheart Director, Regulatory Affairs Southern Nuclear Operating Company CAG/was/cbg
Enclosure:
SNC Response to Request for Additional Information (RAI) cc:
Regional Administrator, Region ll NRR Project Manager - Farley 1 & 2 Senior Resident Inspector - Farley 1 & 2 Alabama - State Health Officer for the Department of Public Health RType: CFA04.054
ENCLOSURE Joseph M. Farley Nuclear Plant - Units 1 and 2 Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specification 3.4.10, Pressurizer Safety Valves to Decrease Low Side Setpoint Tolerance Limiting Condition for Operation (LCO) Value SNC Response to Request for Additional Information
NL-23-0008 Enclosure SNC Response to Request for Additional Information (RAI)
E-1 REQUEST FOR ADDITIONAL INFORMATION (RAI)
By letter dated June 30, 2022 (Agencywide Documents and Access Management System (ADAMS) Accession No. ML22181B145), Southern Nuclear Operating Company (SNC, the licensee) submitted a licensee amendment request (LAR) to propose modifications to the Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2, Technical Specifications (TS) 3.4.10, Pressurizer Safety Valves, regarding the Limiting Condition for Operation (LCO) for pressurizer safety valves (PSVs). The proposed changes would decrease the low side setpoint tolerance value from -1 percent to -2.5 percent for the PSVs.
After reviewing the LAR, the NRC staff requests response to the request for additional information (RAI) given below.
RAI Question #1:
Regulatory Requirements The regulation Title 10 of the Code of Federal Regulation (10 CFR), part 50, Section 36(c)(1) requires that plant TS will include safety limits, limiting safety system settings, and limiting control settings. The regulation 10 CFR 50.36(c)(2)(ii)(C) specifies that a LCO be established for a structure, system, or component (SSC) 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 or presents a challenge to the integrity of a fission product barrier. Farleys PSVs provide, in conjunction with the reactor protection system, overpressure protection for the reactor coolant system.
Licensee LAR Discussion In Section 2.2, Reason for the Proposed Change, the licensee stated:
This change is proposed to reduce an unnecessarily restrictive LCO. FNP [Farley Nuclear Plant] has PSVs manufactured by Crosby. There have been six instances since 2015 where the PSVs were tested and found outside the +/-1% tolerance limits. All out of tolerance test results were outside the low end of the setpoint tolerance (-1%). These as-found results have prompted the generation of licensee event reports (LERs) in accordance with 10 CFR 50.73(a)(2)(i)(B), Any operation or condition prohibited by the plants Technical Specifications... The as-found results did not exceed -3% of the pressure setpoint. Based on the lift pressure meeting the Inservice Test (IST) program requirements, no IST scope expansion testing was needed. Since the as-found result was lower than the allowed value in the TS, the condition did not have an adverse impact on its overpressurization function.
This is within the safety analysis assumptions that are credited for PSVs, and the plant remained bounded by the accident analyses in the Final Safety Analysis Report (FSAR).
Setpoint drift was determined to be the cause of the PSVs lifting low out of tolerance. The PSVs are performing within the design analysis assumptions. Therefore, generating a LER for a PSV that is performing satisfactorily within the design analysis assumptions becomes an unnecessary burden for both the licensee and the NRC.
NL-23-0008 Enclosure SNC Response to Request for Additional Information (RAI)
E-2 RAI #1 In addition to the LAR discussion of the advantages of the proposed low side setpoint tolerance value of -2.5 percent for the Farley PSVs, the licensee is requested to specify whether the evaluations following the six instances of out-of-tolerance test results for the Farley PSVs indicated any material condition concerns with these valves or the performance of their design capabilities. As referenced in the Farley LAR, this information was provided in the LAR submitted by Exelon for the Byron and Braidwood PSV tolerances on June 27, 2003 (ADAMS ML03181034).
SNC Response:
The field service reports for each pressurizer safety valve which failed the as-found test (the subject of the 6 LERs) have been reviewed, and it has been determined that there were no material concerns with the valves that would have prevented them from performing their design functions and that all the valves performed within their design capabilities.
RAI Question #2:
Regulatory Basis The regulation 10 CFR 50, Appendix A, GDC 15, Reactor Coolant System Design, states that the reactor coolant system and associated auxiliary, control, and protection systems are designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.
RAI #2 In Section 3.0 of the enclosure to the letter dated June 30, 2022, under heading Margin Between High Pressure Reactor Trip and PSVs, refer to the following statement:
The calculated uncertainty associated with the Pressurizer Pressure - High RTS [reactor trip system] function was determined to be +28.8 psi [pounds per square inch]. The setpoint uncertainty was determined using a Farley-specific setpoint report utilizing a statistical methodology. The Westinghouse setpoint uncertainty calculation methodology was selected because it was used at many Westinghouse PWRs [pressurized-water reactors], conformed to industry practices such as ISA [International Society of Automation] Standard S67.04, 1987, Setpoints for Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants, and was previously approved by the NRC.
Provide responses to the following:
(a) Briefly describe the analysis for calculating the +28.8 psi uncertainty (including the statistical combination method of uncertainty components) in the nominal high pressurizer pressure reactor trip setting of 2385 psi gauge (psig) for the Pressurizer Pressure - High Reactor Trip System (RTS) function.
NL-23-0008 Enclosure SNC Response to Request for Additional Information (RAI)
E-3 SNC Response:
The basic methodology used is the square root of the sum of the squares (SRSS) technique which has been utilized in other Westinghouse reports. This technique, or others of a similar nature, has been used in WCAP-10395 and WCAP-8567. WCAP-8567 is approved by the NRC noting acceptability of statistical techniques for the application requested. Also, various ANSI, American Nuclear Society, and Instrument Society of America standards approve the use of probabilistic and statistical techniques in determining safety-related setpoints. The basic methodology used in this report is essentially the same as that noted in an ISA paper presented in 1992. This methodology was employed to justify Trip Setpoint and Allowable Value changes in multiple Farley licensing amendments, including Intermediate Range, OTT and OPT reactor trips.
The basic relationship between the uncertainty components and the calculated uncertainty for a channel is noted in the equation below. However, it should be recognized that function specific algorithms may be modified to reflect the plant procedures and the complexity of the function.
CSA = {(PMA)2 + (PEA)2 + (SMTE + SD)2 + (SPE)2 + (STE)2 + (SRA)2 + (RMTE + RD)2 +
(RMTE + RCA)2 + (RRA)2 + (RMTE + RCSA)2 + (RTE)2}1/2 + {(SMTE + SCA)2}1/2 + EA +
BIAS where:
CSA
=
Channel Statistical Allowance PMA
=
Process Measurement Accuracy PEA
=
Primary Element Accuracy SMTE
=
Sensor Measuring & Test Equipment Accuracy SD
=
Sensor Drift SCA
=
Sensor Calibration Accuracy SPE
=
Sensor Pressure Effects STE
=
Sensor Temperature Effects SRA
=
Sensor Reference Accuracy RMTE
=
Rack Measuring & Test Equipment Accuracy RD
=
Rack Drift RCA
=
Rack Calibration Accuracy RRA
=
Rack Reference Accuracy RCSA
=
Rack Comparator Setting Accuracy RTE
=
Rack Temperature Effects EA
=
Environmental Allowance BIAS
=
One directional, known magnitude For the Pressurizer Pressure - High Reactor Trip System (RTS) function, the span of the pressure transmitter was multiplied by the Channel Statistical Allowance (CSA), which is measured in % of span, resulting in the value 28.8 psi.
(b) Regulatory Guide (RG) 1.105, Revision 4 (ADAMS Accession No. ML20330A239), endorses American National Standards Institute (ANSI)/International Society of Automation (ISA)
Standard 67.04.01-2018, Setpoints for Nuclear Safety-Related Instrumentation, for uncertainty calculation. However, according to the above statement, ISA Standard S67.04,
NL-23-0008 Enclosure SNC Response to Request for Additional Information (RAI)
E-4 1987 is used for this calculation. Confirm that the +28.8 psi uncertainty in the Pressurizer Pressure - High RTS function trip setpoint would be obtained by using the NRC endorsed ISA Standard 67.04.01-2018. If not, what would be the uncertainty based on the 2018 ISA standard?
SNC Response:
The statement quoted above from Section 3.0 of the enclosure to the letter dated June 30, 2022, under heading Margin Between High Pressure Reactor Trip and PSVs, was taken from the preface to WCAP-13751, Westinghouse Setpoint Methodology for Protection Systems Farley Nuclear Plant Units 1 and 2 (Model 54F Steam Generators and 2841 MWt NSSS Power), which describes the history and initial basis for the setpoint methodology chosen for Farley Nuclear Plant (FNP). WCAP-13751 was submitted to the NRC for review as part of a License Amendment Request (LAR) requesting Low Feedwater Flow Reactor Trip Elimination and Low-Low Steam Generator Level Setpoint Change (ML20057B439). In the Safety Evaluation Report (SER) provided with Amendments 104 and 97 to Farley Nuclear Plant (FNP) Units 1 and 2, respectively (ML013130715), the NRC stated in the Conclusion that The revised setpoint for the low-low steam generator water level reactor trip and auxiliary feedwater actuation are consistent with the safety limit assumed in the FSAR analysis and are consistent with approved setpoint methodology [emphasis added]. In addition, the revised setpoints remove an over conservatism which contributes to unnecessary reactor trips. On this basis, the staff finds the proposed change in the low-low steam generator water level reactor trip setpoint and allowable value to be acceptable.
Since the Farley setpoint methodology is consistent with approved setpoint methodology, there is no need to re-evaluate the proposed licensing action against RG 1.105, Revision 4, which endorses ANSI/ISA Standard 67.04.01-2018, Setpoints for Nuclear Safety-Related Instrumentation, for uncertainty calculations, and SNC requests that the NRC evaluate this change based on the FNP current licensing basis.
(c) Based on the +28.8 psi uncertainty, the RTS highest trip pressure would be (2385 + 28.8) =
2413.8 psig, which is less by 9.2 psig from the proposed 2423 psig PSV lower limit. If, in response to item (b) above, based on the ISA Standard 67.04.01-2018, the RTS high trip pressure exceeds the proposed PSV lower limit of 2423 psig, the PSV may actuate prior to the RTS trip. In this scenario, provide an evaluation of the impact on the analysis of events/accidents that result on the overpressurization of the reactor coolant system.
SNC Response:
This question is based on the assumption that the uncertainty in the Pressurizer Pressure -
High RTS function trip setpoint was recalculated using ISA Standard 67.04.01-2018 and that the recalculation resulted in the RTS high trip pressure exceeding the proposed PSV lower limit of 2423 psig. As stated in the response to question 2.b. above, since the uncertainty was calculated using the Farley setpoint methodology, which is consistent with approved setpoint methodology, the uncertainty was not recalculated using ISA Standard 67.04.01-2018 and therefore, this question is not applicable.
(d) Define the acronym SAL given in the last sentence of the first paragraph under heading Margin Between High Pressure Reactor Trip and PSVs, Does it stand for Analytical Limit? Explain the meaning of SAL?
NL-23-0008 Enclosure SNC Response to Request for Additional Information (RAI)
E-5 SNC Response:
The acronym SAL given in the last sentence of the first paragraph under heading Margin Between High Pressure Reactor Trip and PSVs, stands for Safety Analysis Limit. The SAL is the analytical limit assumed in the plant safety analyses and/or specified in the plant design basis. The SAL is the limiting parameter value in the safety and transient analysis at which a reactor trip or ESF actuation function is assumed to be initiated.