ML24024A137
| ML24024A137 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 01/24/2024 |
| From: | James Kim Plant Licensing Branch 1 |
| To: | Hasanat A Exelon FitzPatrick |
| References | |
| L-2023-LLA-0103 | |
| Download: ML24024A137 (5) | |
Text
From:
James Kim Sent:
Wednesday, January 24, 2024 11:43 AM To:
Hasanat, Abul M:(Constellation Nuclear)
Cc:
Hawes, Mark:(Constellation Nuclear); Theo Edwards
Subject:
Final SNSB RAI regarding FitzPatrick Amendment to Modify Safety Relief Valves Setpoint Lower Tolerance (EPID: L-2023-LLA-0103)
Attachments:
Fitzpatrick SRV_SNSB RAIs.docx
SUBJECT:
FitzPatrick - Final SNSB RAI regarding Amendment to Modify Safety Relief Valves Setpoint Lower Tolerance (EPID: L-2023-LLA-0103)
Mr. Hasnat, By letter dated July 28, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession ML23209A003), Constellation Energy Generation, LLC (CEG) submitted an amendment to revise the James A. FitzPatrick Nuclear Power Plant (FitzPatrick) Technical Specifications 3.4, Reactor Coolant System (RCS), Section 3.4.3, Safety/Relief Valves (S/RVs). Specifically, CEG proposes a new safety function lift setpoint lower tolerance for the S/RVs as delineated in SR 3.4.3.1. The proposed change would revise the lower setpoint tolerance from -3 percent to -5 percent.
The NRC staff has determined that additional information is needed to complete its review of the amendment. On January 11, 2023, the NRC staff sent FitzPatrick the draft Request for Additional Information (RAI) from the Nuclear Systems Performance Branch (DSS/SNSB). On January 24, 2024, the RAI clarification call was held between the NRC and FitzPatrick staff and the licensee agreed to provide the RAI responses by February 29, 2024. A publicly available version of this final RAI (attached) will be placed in the NRCs ADAMS.
James Kim Project Manager - FitzPatrick NRR/DORL/LPL1 301-415-4125
Hearing Identifier:
NRR_DRMA Email Number:
2378 Mail Envelope Properties (DM6PR09MB47113E302B753BF80BC1235BE47B2)
Subject:
Final SNSB RAI regarding FitzPatrick Amendment to Modify Safety Relief Valves Setpoint Lower Tolerance (EPID L-2023-LLA-0103)
Sent Date:
1/24/2024 11:42:43 AM Received Date:
1/24/2024 11:42:00 AM From:
James Kim Created By:
James.Kim@nrc.gov Recipients:
"Hawes, Mark:(Constellation Nuclear)" <Mark.Hawes@constellation.com>
Tracking Status: None "Theo Edwards" <Theo.Edwards@nrc.gov>
Tracking Status: None "Hasanat, Abul M:(Constellation Nuclear)" <Abul.Hasanat@constellation.com>
Tracking Status: None Post Office:
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REQUEST FOR ADDITIONAL INFORMATION BY NUCLEAR SYSTEMS PERFORMANCE BRANCH JAMES A. FITZPATRICK NUCLEAR POWER PLANT CHANGE IN TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT 3.4.3.1 DOCKET NO. 50-333 EPID: L-2023-LLA-0103 INTRODUCTION By application dated July 28, 2023, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23209A003) (Reference 1), Constellation Energy Generation, LLC (CEG, the licensee) submitted a license amendment request (LAR) requesting change to the Technical Specifications (TS) for the James A. FitzPatrick Nuclear Power Plant (JAF) Renewed Facility Operating License DPR-59. The proposed change would revise the safety function lift setpoint tolerances for the Safety/Relief Valves (S/RVs) that are listed in Surveillance Requirement (SR) 3.4.3.1 of the Technical Specifications (TS). This change would be limited to the lower tolerances and would not affect the upper limits. The tolerance band for these valves would be changed from +/-3% to +3% or -5% of the setpoint (1145 psig +34.3 or -57.2 psig).
After reviewing the LAR (Reference 1), the Nuclear Systems Performance Branch (SNSB) staff requests response to the request for additional information (RAI) given below.
SNSB-RAI 1 Regulatory Basis:
10 CFR 50, Appendix A, GDC Criterion 16, Containment design. Reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.
10 CFR 50, Appendix A, GDC Criterion 50, Containment design basis. The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident. This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and as required by § 50.44 energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergency core cooling functioning, (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters.
RAI:
Refer to General Electric (GE) proprietary report NEDE-22223, "Low-Low Set Logic and Lower MSIV Water Level Trip for BWRs with Mark I Containment, dated September 1982 (Reference
- 2) which proposed design modifications for BWR/2-4, with Mark I containments to limit containment loading following subsequent S/RV actuations. The intent of these modifications was to reduce the discharge loads on the containment and suppression pool structures resulting from subsequent S/RV actuations during a transient.
The proposed modifications consisted of a low-low set (LLS) relief logic system, and a lower main steam isolation valve (MSIV) water level trip. The LLS is an automatic S/RV control system while the lower MSIV water level trip modification is applicable to BWR/4 plants only.
As JAF is a BWR/4, Mark I containment plant:
(a) Did JAF implement the recommended modifications for BWR/4 with Mark I containments as described in NEDE-22223?
(b) The modifications (if implemented) were developed with the lower setpoint at -3% of the value. The submitted LAR proposes a change to the lower setpoint for the S/RVs from -3%
to -5%. Please describe the impact this setpoint change has on containment loads during subsequent S/RV actuations during transients as reported in NEDE-22223.
(c) If JAF has not implemented the recommended modifications, provide reasons for not implementing the recommended changes and describe the impact the proposed setpoint tolerance change has on containment loading during subsequent S/RV actuations during the applicable transients.
SNSB-RAI 2 Regulatory Basis:
10 CFR 50, Appendix A GDC Criterion 13, Instrumentation and control. Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
10 CFR 50, Appendix A, GDC 29, Protection against anticipated operational occurrences,
The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.
RAI:
Reference 1 discusses operating margin in relation to nominal operating pressure in Section 3.3 of Attachment 1, with the statement:
The purpose of the lower setpoint tolerance is to ensure sufficient margin exists between the normal operating pressure of the system and the point at which the S/RVs actuate in the overpressure safety mode. The nominal operating pressure of the reactor pressure vessel power is 1040 psig. A lower setpoint tolerance value of -5%, applied to the S/RV set pressure (1145 psig) would allow it to lift at 1087.8 psig.
(a) What is the tolerance (uncertainty) range of the reactor instrumentation that provides a nominal reading of 1040 psig?
(b) Please describe the impact of the proposed change and justify sufficient margin remains between the upper tolerance limit of the normal nominal operating pressure and the proposed lowest S/RV opening pressure of 1087.8 psig so that the S/RVs would not actuate during normal plant operation.
Similarly, in TS Table 3.3.1.1-1, the reactor protection system (RPS) Function 3 has the reactor pressure allowable value 1080 psig.
(c) Provide the RPS scram setpoint during an overpressure transient and the margin between the scram setpoint pressure and the proposed lowest S/RV opening pressure of 1087.8 psig. Justify the margin is sufficient so that the S/RV would not operate prior to the RPS actuation.
REFERENCES
- 1. Letter from CEG to NRC, License Amendment Request to Modify Technical Specification Surveillance Requirement (SR) 3.4.3.1 Safety Relief Valves (S/RVs) Setpoint Lower Tolerance, July 28, 2023, ADAMS Accession No. ML23209A003.
- 2. NEDE-22223, Low-Low Set Logic and Lower MSIV Water Level Trip for BWRs with Mark I Containment, dated September 1982. ADAMS Accession No. ML19262G927