ML24136A116
| ML24136A116 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/26/2024 |
| From: | James Kim Plant Licensing Branch 1 |
| To: | Rhoades D Constellation Energy Generation |
| Poole J | |
| References | |
| EPID L-2023-LLA-0103 | |
| Download: ML24136A116 (15) | |
Text
June 26, 2024
David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT NO. 355 RE: REVISE TECHNICAL SPECIFICATIONS SECTION 3.4.3.1, SAFETY RELIEF VALVES SETPOINT LOWER TOLERANCE (EPID L-2023-LLA-0103)
Dear David Rhoades:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 355 to Renewed Facility Operating License No. DPR-59 for the James A. FitzPatrick Nuclear Power Plant. The amendment consists of changes to the technical specifications in response to your application dated July 28, 2023, as supplemented by letter dated February 29, 2024.
The amendment revises the Technical Specifications 3.4, Reactor Coolant System (RCS),
Section 3.4.3, Safety/Relief Valves (S/RVs). Specifically, Constellation Energy Generation, LLC proposes a new safety function lift setpoint lower tolerance for the S/RVs as delineated in SR 3.4.3.1. The proposed change revises the lower setpoint tolerance from -3 percent to -5 percent.
D. Rhoades
A copy of our related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
James S. Kim, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Docket No. 50-333
Enclosures:
- 1. Amendment No. 355 to DPR-59
- 2. Safety Evaluation
cc: Listserv CONSTELLATION FITZPATRICK, LLC
AND
CONSTELLATION ENERGY GENERATION, LLC
DOCKET NO. 50-333
JAMES A. FITZPATRICK NUCLEAR POWER PLANT
AMENDMENT TO RENEWED FA CILITY OPERATING LICENSE
Amendment No. 355 Renewed Facility Operating License No. DPR-59
- 1. The U.S. Nuclear Regulatory Commission has found that:
A. The application for amendment by Exelon Generation Company, LLC, dated July 28, 2023, as supplemented by letter dated February 29, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions rules and regulations set forth in 10 CFR Chapter I;
B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;
C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations;
D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and
E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
Enclosure 1
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-59 is hereby amended to read as follows:
(2) Technical Specifications
The Technical Specifications contained in Appendix A, as revised through Amendment No. 355, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of its issuance and shall be implemented within 45 days.
FOR THE NUCLEAR REGULATORY COMMISSION
Hipólito González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications
Date of Issuance: June 26, 2024
ATTACHMENT TO LICENSE AMENDMENT NO. 355
JAMES A. FITZPATRICK NUCLEAR POWER PLANT
RENEWED FACILITY OPERATING LICENSE NO. DPR-59
DOCKET NO. 50-333
Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Page Insert Page Page 3 Page 3
Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contain marginal lines indicating the areas of change.
Remove Page Insert Page 3.4.3-2 3.4.3-2
(3) Constellation Energy Generation, LLC, pursuant to the Act a nd 10 CFR Parts 30, 40, and 70, to receive, possess, and use, at any time, any byproduct, source, and special nuclear material as sealed neutr on sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required;
(4) Constellation Energy Generation, LLC, pursuant to the Act a nd 10 CFR Parts 30, 40, and 70, to receive, possess, and use, at any time, any byproduct, source, and special nuclear material without restric tion to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus, components or tools.
(5) Constellation Energy Generation, LLC, pursuant to the Act a nd 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct a nd special nuclear materials as may be produced by the operation of the facility.
C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:
Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Se ctions 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional co nditions specified or incorporated below:
(1) Maximum Power Level Constellation Energy Generation, LLC is authorized to operate t he facility at steady state reactor core power levels not in excess of 2536 megawatts (thermal).
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revise d through Amendment No. 355, are hereby incorporated in the renewed opera ting license. The licensee shall operate the facility in accordance with the Technical Specifications.
Amendment 355 Renewed License No. DPR-59 S/RVs 3.4.3
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
SR 3.4.3.1 Verify the safety function lift setpoint of the In accordance required S/RVs is 1145 + 34.3 or - 57.2 psig. with the Following testing, lift settings shall be within INSERVICE
+/- 1%. TESTING PROGRAM
SR 3.4.3.2 Verify each required S/RV is capable of being In accordance opened. with the INSERVICE TESTING PROGRAM
JAFNPP 3.4.3-2 Amendment 355 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
RELATED TO AMENDMENT NO. 355
CONSTELLATION FITZPATRICK, LLC
CONSTELLATION ENERGY GENERATION, LLC
DOCKET NO. 50-333
JAMES A. FITZPATRICK NUCLEAR POWER PLANT
TO RENEWED FACILITY OPERATING LICENSE NO. DPR-59
1.0 INTRODUCTION
By letter dated July 28, 2023 (Agencywide Documents Access and Management System Accession No. ML23209A003) as supplemented by letter dated February 29, 2024 (ML24060A032), the Constellation Energy Generat ion, LLC (Constellation, the licensee),
submitted a license amendment request to revise James A. Fitzpatrick Nuclear Power Plant (FitzPatrick) Technical Specifications (TSs). 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 TS. This change would be limited to the lower tolerances and would not affect the upper limits. The setpoint tolerance band for these valves would be changed from +/-3% of the setpoint (1145 psig +/- 3%) to +3% or -5% of the setpoint (1145 psig +34.3 or -57.2 psig).
The supplemental letter dated February 29, 2024, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on October 31, 2023 (88 FR 74530).
2.0 REGULATORY EVALUATION
2.1 Description of System
The pressure relief system consists of 11 Target Rock three stage S/RVs, model 0867F-001, located in the drywell on the main steam lines be fore the first main steam isolation valves. The pressure relief system uses S/RVs to protect the reactor coolant pressure boundary (RCPB) from damage due to overpressure by discharging steam from the reactor coolant system to the suppression pool.
Enclosure 2
Each valve is automatically actuated by steam pressure in excess of its setpoint or manually using remote switches. Seven of these valves make up the Automatic Depressurization System (ADS), which further reduces reactor pressure in the event of a small break loss of coolant accident (LOCA) so that the low-pressure coolant injection and core spray systems can operate to protect the fuel barrier.
2.2 Proposed Change
The licensee is proposing to change the tolerance band for S/RVs as measured in TS SR 3.4.3.1 from its current +/-3% tolerance around the nominal lift setpoint of 1145 psig (1179.3 psig - 1110.7 psig) to a +3 or -5% tolerance (1179.3 psig - 1087.8 psig). The licensee stated that in the past 16 tests, there have been five instances in which the S/RVs tested as-found setpoints were outside the low end (-3%) tolerance limit of the nominal setpoint, which resulted in those valves declared inoperable. The test results also indicated that the as-found lift settings were all within the proposed new lower limit of -5% of the nominal 1145 psig pressure setting.
2.3 Regulatory Requirements
The applicable Title 10 of the Code of Federal Regulations (10 CFR) Part 50 requirements are as follows:
10 CFR 50.36(c)(1) requires that plant TSs will include safety limits, limiting safety system settings, and limiting control settings.
10 CFR 50.36(c)(2)(ii)(C) specifies that a limiting condition for operation be established for 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 or presents a challenge to the integrity of a fission product barrier.
NUREG-0800, Standard Review Plan (SRP), 5.2. 2, Overpressure Protection,Section II states the acceptance criteria are based on meeting the relevant requirements of the following Commission regulations:
General Design Criteria (GDC) 15, as it relates to designing the RCPB and associated auxiliary, control, and protection systems with sufficient margin to assure that the design conditions of the RCPB are not exceeded during any condition of normal operation, including anticipated operational occurrences (AOOs).
GDC 31, as it relates to designing the RCPB with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions the boundary behaves in a nonbrittle manner and the probability of rapidly propagating fractures is minimized.
NUREG-0800, Standard Review Plan 5.2.2, in Section II, under heading SRP Acceptance Criteria, in Section 2.B, requires the design of safety valves should have sufficient capacity to limit the pressure to less than 110 percent of the RCPB design pressure during the most severe AOO with reactor scram, as specifi ed by ASME Code Article NB-7000. Sufficient available margin should account for uncertainties in the design and operation of the plant.
3.0 TECHNICAL EVALUATION
The licensees and NRC staff evaluation of the events that model the S/RV setpoint and may be affected by the proposed change is given below.
3.1 Thermal Limits
The licensee evaluated the effect of changing the lower S/RV setpoint level on the operating limit minimum critical power ratios (MCPRs), off-rated MCPRs, and the off-rated linear heat generation rates (LHGRs) and concluded there was no impact on the thermal limits events analyses. Thermal limits AOOs credit the relief mode of the S/RV and the analysis uses setpoints greater than the allowable upper tolerance, which will be unaffected by the proposed changes to the lower tolerance.
The NRC staff reviewed the licensees assumpti ons and explanations of the effect of the proposed change and concludes that the thermal limits AOO analyses and MCPR and LHGR limits remain valid because the AOOs credit the relief mode of the S/RVs which does not involve their setpoint.
3.2 ASME Overpressure
The Final Safety Analysis Report 2023 for FitzPatrick, Section 14.5 (ML23215A212) discusses the analyses of abnormal operational transients and reactor vessel overpressure. These analyses state the limiting event is a main steam isolation valve (MSIV) closure, terminated by a high neutron flux reactor scram, meaning that there was a failure of the direct scram from MSIV position. The licensee states that these analyses are based on the S/RV upper tolerance setting, which will be unaffected by the proposed changes to the lower tolerance.
The NRC staff concludes that the proposed change to the S/RVs lower tolerance setting will have no impact on the ASME overpressure analyses because it is based on the upper tolerance setting as an input.
3.3 LOCA Analysis
The licensee states that there is no adverse impact to the LOCA analysis because of the proposed changes. The S/RVs provide a mechanism for reactor depressurization during a small break LOCA through the ADS function, which is not based on high pressure safety setpoints and will be unaffected by the proposed change in the S/RVs lower tolerance setting.
For a large break LOCA, the break itself depressurizes the reactor and the S/RVs are not needed to mitigate the accident.
The NRC staff concludes that the proposed change to the S/RVs lower tolerance setting will have no impact on the LOCA analyses because in a small break LOCA, the S/RVs are operated by the ADS in their relief mode, while in a large break LOCA, the reactor vessel is depressurized by the break so the S/RVs will not operate either in their relief or safety mode.
3.4 High Pressure System Performance
This refers to the reactor core isolation cooling system, high pressure coolant injection (HPCI) system, and standby liquid control system performance to inject makeup water into the reactor vessel during accident scenarios where the reactor pressure remains high. The licensee states
that the potential effect on these high-pressure systems because of a change to the S/RV setpoint would be the maximum reactor pressure at which they are required to deliver flow to the reactor. As discussed in Section 3.2, the licensee states that these analyses are based on the S/RV upper tolerance setting, which will be unaffected by the proposed changes to the lower tolerance.
The NRC staff concludes that the proposed change to the lower tolerance setting will have no impact on the high pressure systems performance analyses because they are based on the upper tolerance setting as an input.
3.5 Anticipated Transient Without Scram (ATWS) Mitigation (Peak Suppression Pool Temperature)
The licensee discusses this case as being the same as the overpressure analysis discussed in Section 3.2 above, with the initial conditions of an ATWS. As discussed in Section 3.2, the licensee states that these analyses are based on the S/RV upper tolerance setting, which will be unaffected by the proposed changes to the lower tolerance.
The NRC staff concludes that the proposed change to the lower S/RVs tolerance setting will have no impact on the ATWS event analyses because it is based on the upper tolerance setting as an input.
3.6 Containment Analyses
The licensee states that the existing LOCA containment pressure and temperature response analyses remains bounded with the proposed S/RV setpoint changes. Their assertion is The flow rate through the S/RVs is directly proportional to the pressure differential across them.
Since allowing lower pressure differences at the minimum tolerance end does not affect the limiting analyses at the maximum tolerance end of the pressure range, there is no impact on the containment systems, structures, or components ability to manage energy release during S/RV actuation. The NRC staff had further questions a bout this assertion which are discussed in Section 3.7.
3.7 S/RV Discharge Loads Resulting from Subsequent S/RV Actuation
In a request for additional information, the NRC staff referred to General Electric proprietary report NEDE-22223, Low-Low Set [LLS] Logic and Lower MSIV Water Level Trip for boiling water reactors (BWRs) with Mark I Containment, dated September 1982 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 as recommended in NEDC-22223 was to reduce the discharge loads on the containment and suppression pool structures resulting from subsequent S/RV actuations during a transient. The NRC staff requested clarification on whether there were potential impacts of the lower setpoint change to the licensees analysis of subsequent S/RV actuations. In the response letter dated February 29, 2024, the licensee referred to a previous analysis, Mark I Containment Program Plant Unique Analysis Report, JPN-83-88, dated October 11, 1983, ML20078H509, which includes a discussion of historical modifications developed to reduce hydrodynamic loading in the containment to within analyzed values. This analysis was performed in response to NEDE-22223. The licensee stated that the LLS logic modification was not performed as there is no S/RV opening logic associated with the low allowable value of the S/RV setpoints. The licensee further stated that the current S/RV dynamic load analysis, conservatively use an upper bound
of the allowable value (1145 psig + 3%) as the assumed S/RV lift pressure. The transient evaluations consider the second S/RV actuation to occur at the maximum reflood and demonstrated that allowable stresses due to loads in the S/RV discharge piping and containment are not exceeded. The lowering of the allowable value is bounded by existing analysis and does not present an adverse change to containment loads.
In letter dated February 29, 2024, regarding the implementation of recommendations in NEDC-22223, the licensee stated:
FitzPatrick did not implement all recommended modifications of NEDE-22223. As discussed above, FitzPatrick met Mark I Containment Long Term Program stress requirements absent the implementation of LLS by meeting stress allowable criteria for subsequent actuations at the most limiting timing.
During the review of NEDE-22223, it was identified that BWR Owners group provided recommendations for proposed actions and modifications to meet the objectives of NUREG-0737 (ML051400209), Item II.K.3.16. These actions and modifications were designed to reduce subsequent actuations for plant transients, reactor isolations, and improve overall S/RV performance. The recommendations included:
- 1) LLS Relief Logic System or Equivalent Manual Actions
- 2) Lower the reactor pressure vessel water level isolation setpoint for main steam isolation valve closure from Level 2 to Level 1
FitzPatricks response was that equivalent operator manual actions as implemented by emergency procedures adequately reduced the challenges and failures of relief valves in coordination with preventative maintenance practices and achieves the objectives of NUREG-0737, Item II.K.3.16.
These equivalent manual actions remain in place. Current emergency operating procedures for high reactor pressure (greater than 1080 psig) prompt entry into emergency operating procedures for Hot Reactor Pressure Vessel Control, which directs operators to stabilize pressure below 1080 psig, which is well below the lifting setpoint for S/RV actuation.
Subsequent to this response, FitzPatri ck implemented Amendment No.103 (ML010610096) to lower the MSIV setpoint to Level 1 as recommended.
The proposed change to the allowable value is bounded by the existing analyses.
The NRC staff concludes that even though the recommendations in NEDC-22223 were not implemented in FitzPatrick, the NRC staff finds it acceptable because the licensee adequately justified that the proposed change to the S/RV setpoint lower tolerance setting will have no impact on the S/RV discharge piping and containment loads, and therefore the current analysis remains bounded.
3.8 Discussion on Operating Margin
The licensee states that the purpose for the lower setpoint tolerance is to ensure that sufficient margin exists between the normal operating pressure of the system and the point where the S/RVs actuate.
Current Technical Proposed Technical Specifications Specifications Nominal Operating 1040 psig 1040 psig Pressure Lower S/RV setpoint 1110.70 psig 1087.8 psig Margin 70.7 psig 47.8 psig
The licensee states that the proposed change is for testing acceptance criteria only and all valves are returned to service with setpoints at +/- 1% tolerance.
The NRC staff requested clarification on how in strument accuracy impacts the operating margin, both for normal operating conditions and transient conditions. In a letter dated February 29, 2024, the licensee states that for normal operating conditions, the total instrument uncertainty (calculated from the total loop of uncertainty of the Feedwater Narrow Range Pressure Control Transmitter and associated recorder that provides narrow range reactor pressure indication) results in 40.25 psig of margin, compared to the technical specification margin, nominally calculated as 47.8 psig. The licensee states there is similar margin available with the wide range pressure instrumentation, associated with the high reactor pressure alarm (setpoint of 1051 psig) that draws from inputs from the Reactor Water Level Feedwater Control Compensating Pressure Transmitters. This recorder and associated alarm have a total loop uncertainty of 19.2 psi. Therefore, the reactor high pressure alarm actuates prior to 1070.2 psig, leaving 17.6 psi of margin to the proposed low allowable value for the S/RV mechanical lift.
For transient conditions, overpressure conditions are mitigated by the Reactor Pressure System High Pressure Scram, with a Technical Specifications Allowable Value of 1080 psig.
To ensure the allowable value is not exceeded, the licensee states they have set the RPS high pressure scram setpoint at 1062 psig, which accounts for 14.25 psig loop uncertainty and still provides 11.55 psig of margin between the proposed lowered allowable setpoint.
The NRC staff concludes that sufficient margin remains to ensure the S/RV system operates as designed.
3.9 Technical Conclusion
The NRC staff concludes that the proposed SR 3.4.3.1 change in the S/RV setpoint from its current +/-3% tolerance around the nominal lift setpoint of 1145 psig (1179.3 psig - 1110.7 psig) to a +3 or -5% tolerance (1179.3 psig - 1087.8 psig) is acceptable based on the following:
The proposed change to the lower tolerance setting will not impact the thermal limits analyses.
The proposed change to the lower tolerance setting will have no impact on the ASME overpressure analysis, HPCI performance, or ATWS mitigation which is based on the upper tolerance setting as an input.
The proposed change to the lower tolerance setting will have no impact on the LOCA analyses due to the way the S/RVs function in a small or large break LOCA.
The existing containment analysis remains valid with the proposed changes.
Sufficient operating margin remains during normal and transient conditions.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the New York State official was notified of the proposed issuance of the amendment on May 14, 2024. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (October 31, 2023; 88 FR 74530). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that 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.
Principal Contributor: J. Ambrosini
ML24136A116 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DSS/SNSB/BC NRR/DSS/STSB/BC (A)
NAME JKim KZeleznock PSahd SMehta DATE 05/14/2024 05/16/2024 04/25/2024 05/23/2024 OFFICE OGC - NLO NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME IMurphy HGonzález JKim DATE 06/13/2024 06/26/2024 06/26/2024