CNL-25-039, Response to Request for Additional Information Regarding License Amendment Request to Adopt TSTF-576
| ML25069A477 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/10/2025 |
| From: | Hulvey K Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| CNL-25-039, EPID L-2024-LLA-0163 | |
| Download: ML25069A477 (1) | |
Text
10 CFR 50.90 1101 Market Street, Chattanooga, Tennessee 37402 CNL-25-039 March 10, 2025 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260, and 50-296
Subject:
Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Response to Request for Additional Information Regarding License Amendment Request to Adopt TSTF-576 (EPID L-2024-LLA-0163)
References:
- 1. TVA Letter to NRC, CNL-24-049, "Browns Ferry Nuclear Plant, Units 1, 2, and 3 - License Amendment Request for Adoption of TSTF-576 to Revise Safety/Relief Valve Requirements," dated December 9, 2024 (ML24344A034) 2.
NRC Electronic Mail to TVA, Request for Additional Information - TVA LAR to Revise Browns Ferry Technical Specification 3.4.3 by Adopting TSTF-576 (EPID L-2024-LLA-0163), dated February 10, 2025 (ML25042A041)
In Reference 1, Tennessee Valley Authority (TVA) requested an amendment to Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 for Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, respectively. The request contained proposed changes to BFN Technical Specification (TS) 3.4.3, Safety/Relief Valves (S/RVs), by adopting Technical Specification Task Force (TSTF) Traveler TSTF-576, Revision 3, Revise Safety/Relief Valve Requirements.
In Reference 2, the Nuclear Regulatory Commission (NRC) issued a request for additional information (RAI) and requested that TVA provide a response by March 12, 2025. to this submittal provides the TVA response to the RAI.
In response to NRC RAI-3, Enclosure 2 to this submittal provides revised BFN Units 1, 2, and 3 TS pages marked up to show the proposed changes. The TS pages in Enclosure 2 supersede those provided in Attachment 1 to the enclosure of Reference 1, with changes in blue text.
U.S. Nuclear Regulatory Commission CNL-25-039 Page 2 March 1, 2025 In response to NRC RAI-4 and RAI-5, Enclosure 3 to this submittal provides revised BFN Unit 1 TS Bases pages marked up to show the proposed changes (BFN Units 2 and 3 TS Bases changes will be similar). Changes to the existing TS Bases pages are provided for information only and will be implemented under the TS Bases Control Program. The TS Bases in supersede those provided in Attachment 2 to the enclosure of Reference 1, with changes in blue text.
In response to NRC RAI-1 and RAI-2, Enclosure 4 to this submittal provides a revised sample Core Operating Limits Report (COLR) page for BFN Unit 1. Changes to the existing BFN COLR pages are provided for information only. The sample COLR page in Enclosure 4 supersedes that provided in Attachment 3 to the enclosure of Reference 1.
This submittal does not change the no significant hazards consideration of the environmental consideration contained in Reference 1. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.91, Notice for public comment; State consultation, a copy of this supplement is being provided to the designated Alabama office.
There are no new regulatory commitments associated with this submittal. Please address any questions regarding this request to Amber V. Aboulfaida, Senior Manager, Fleet Licensing, at avaboulfaida@tva.gov.
I declare under penalty of perjury that the foregoing is true and correct. Executed on this 1th day of March 2025.
Respectfully, Kimberly D. Hulvey General Manager, Nuclear Regulatory Affairs & Emergency Preparedness
Enclosure:
1.
TVA Response to NRC Request for Additional Information 2.
Revised TS Changes (Markups) for BFN Units 1, 2, and 3 3.
Revised TS Bases Changes (Markups) for BFN Unit 1 (For Information Only) 4.
Example Updated COLR for BFN Unit 1 (For Information Only) cc (Enclosure):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant NRC Project Manager - Browns Ferry Nuclear Plant State Health Officer, Alabama Department of Public Health Digitally signed by Edmondson, Carla Date: 2025.03.10 09:58:45 -04'00' CNL-25-039 E1 - 1 of 4 TVA Response to NRC Request for Additional Information Introduction By letter dated December 9, 2024 (Agencywide Documents Access and Management System Accession No. ML24344A034), the Tennessee Valley Authority (TVA) submitted a license amendment request (LAR) for the Browns Ferry Nuclear Plant, Units 1, 2, and 3 (Browns Ferry or BFN). Specifically, the proposed amendments would revise Browns Ferry Technical Specification (TS) 3.4.3, Safety/Relief Valves (S/RVs), by adopting Technical Specification Task Force (TSTF) Traveler TSTF-576, Revision 3, Revise Safety/Relief Valve Requirements.
The U.S. Nuclear Regulatory Commission (NRC) is reviewing the LAR, and has identified where additional information is needed to complete its review. The requested information is identified below.
Regulatory Basis The regulation at paragraph 50.36(c)(2)(ii)(C) of Title 10 of the Code of Federal Regulations (10 CFR) requires the establishment of a limiting condition for operation (LCO) 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 of or presents a challenge to the integrity of a fission product barrier.
The regulation at 10 CFR 50.36(c)(3) addresses surveillance requirements (SRs) which are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCO will be met.
Requests
RAI-1
The proposed revision to the Browns Ferry, Unit 1, Core Operating Limits Report (COLR),
shown in Attachment 3 to the Enclosure to the LAR (provided for information only), states, in part:
The as-found OPS lift pressures of the required S/RVs are within +/- 3% of the limits specified below:
Number of S/RVs As-Found Lift Pressure Limit (psig) 4 1135 4
1145 5
1155 Traveler TSTF-576 removed the lower tolerance (i.e., the -3%), however, it is included as part of the sample COLR page but is not marked as a variation.
CNL-25-039 E1 - 2 of 4 The proposed SR 3.4.3.2 requires that the as-found Overpressure Protection System (OPS) lift pressures of the required S/RVs be within the limits specified in the COLR.
Clarify which pressure limits are intended to be used in SR 3.4.3.2 and provide an updated sample COLR page, as appropriate.
TVA Response The lower tolerance of the S/RVs is not a part of the BFN safety analysis of the as-found OPS.
The pressure limits intended to be used in SR 3.4.3.2 include the higher tolerance of the required 12 S/RVs currently credited in OPS safety analysis. Accordingly, an updated sample COLR page has been provided in Enclosure 4 of this RAI response and supersedes that provided in Attachment 3 to the enclosure of the LAR.
RAI-2
The Applicable Safety Analyses portion of the proposed TS 3.4.3 Bases states that the overpressure analyses assume 12 S/RVs operate in the safety mode of operation. However, the sample COLR page shows that 13 S/RVs are required. The values in the COLR should be consistent with, or more conservative, than what is assumed in the safety analysis.
Clarify how many S/RVs are required to operate in the safety mode and update the Bases or COLR, as appropriate, to resolve this apparent inconsistency.
TVA Response As indicated in the response to RAI-1, BFN safety analysis currently requires 12 S/RVs to operate in the safety mode. This has been addressed in the updated sample COLR page provided in Enclosure 4.
RAI-3
Traveler TSTF-576, Revision 3, states that plant-specific amendments to adopt the proposed change will add the OPS specification to paragraph a of the COLR TS and notes that this addition should be consistent in format with the existing list of specifications referenced there.
Section 3.1 of the Enclosure to the LAR states [t]he COLR specification is revised to reference the OPS specification; however, there were no proposed changes to TS 5.6.5.a to add the reference to proposed TS 3.4.3, Overpressure Protection System (OPS).
Explain why a proposed change to TS 5.6.5.a is not needed, or propose a revision to TS 5.6.5.a to reflect the reference to TS 3.4.3.
TVA Response A proposed revision to BFN TS 5.6.5.a to include the as-found OPS lift pressures for TS 3.4.3 is provided in the updated TS markups in Enclosure 2 of this RAI response. Additionally, the revised TS pages contain minor editorial changes that correct inadvertent deviations from the TSTF traveler.
CNL-25-039 E1 - 3 of 4
RAI-4
The proposed SR 3.4.3.1 requires the required as-left S/RV settings to be within +/-1% of the nominal setpoint of the valve. The SR lists all 13 installed valves. The applicable TS Bases markup states that 12 valves are required to perform the OPS function.
The NRC staff determined that any S/RV that is not set per SR 3.4.3.1 cannot be credited in the as-found test group (SR 3.4.3.2, based on the COLR valve settings list) because the valve would not be verified to be operable based on the as-left setting.
LCO 3.0.1 requires that the S/RVs credited for the OPS must be operable in all modes and specified conditions of the applicability. The lack of an acceptable as-left test would make the S/RV unable to contribute to OPS operability.
Explain how the operators will ensure that only the S/RVs that are determined to be operable by as-left testing per SR 3.4.3.1 are credited when performing SR 3.4.3.2.
TVA Response A variation to TSTF-576 is proposed to Bases SR 3.4.3.2 to ensure that only S/RVs determined to be operable in SR 3.4.3.1 can be credited in SR 3.4.3.2. Section 2.2 of the LAR is supplemented as follows:
The revised Bases SR 3.4.3.2 contains language clarifying that only as-left valves credited for SR 3.4.3.1 can be credited for SR 3.4.3.2. This wording is additional to the STS on which TSTF-576 was based but reconciles that 12 of 13 S/RVs are required to be operable in all modes and conditions of applicability.
Accordingly, proposed revisions to TS Bases 3.4.3 are provided in Enclosure 3 of this RAI response that reflect this change to Bases SR 3.4.3.2.
RAI-5
Section 2.2, Optional Changes and Variations, of the LAR states that the LCO section of the Bases 3.4.3 contains a bracketed inclusion of relief mode for overpressure protection and it is applicable as all S/RVs at Browns Ferry are capable of safety and relief mode for overpressure protection.
This appears to be contradictory to the proposed revision for the Background section of TS Bases 3.4.3, which states that The safety mode is credited for overpressure protection, and S/RVs operating in relief mode are not credited for overpressure protection.
Additionally, the proposed revision for the LCO section of TS Bases 3.4.3 (bottom of Bases page B 3.4-18) states:
The OPS is OPERABLE when it can ensure that the ASME Code limit on peak reactor pressure, as stated in Safety Limit 2.1.2, will be protected using the safety and relief modes [emphasis added] of the S/RVs. The OPERABILITY of the OPS is only dependent on the ability to relieve excess pressure and maintain reactor pressure below Safety Limit 2.1.2, and may credit less than the full complement of installed S/RVs.
CNL-25-039 E1 - 4 of 4 Explain these apparent contradictions regarding the safety mode of the OPS, or propose appropriate revisions to the TS Bases and/or variations to Traveler TSTF-576, Revision 3.
TVA Response Only the safety mode of the S/RVs is credited with overpressure protection, so the bracketed inclusion of relief mode is not applicable to BFN. The statement in Section 2.2 of the LAR was erroneous and should instead read as follows:
The LCO section of Bases 3.4.3 contains a bracketed inclusion of relief mode for overpressure protection, and this is applicable as all S/RVs at BFN are capable of safety and relief mode for overpressure protection but this is not applicable at BFN because the safety mode of the S/RVs is credited with overpressure protection.
Accordingly, a proposed revision to the Limiting Condition for Operation section of TS Bases 3.4.3 is provided in Enclosure 3 of this RAI response that resolves these contradictions.
CNL-25-039 Revised Technical Specifications Changes (Markups) for BFN Units 1, 2, and 3 (9 Pages)
S/RVs 3.4.3 BFN-UNIT 1 3.4-7 Amendment No. 234 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety/Relief Valves (S/RVs)
LCO 3.4.3 The safety function of 12 S/RVs shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required S/RVs inoperable.
A.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND A.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OPS Overpressure Protection System (OPS)
OPS OPS inoperable.
S/RVs 3.4.3 BFN-UNIT 1 3.4-8 Amendment No. 234, 263, 269, 301, 315, 333 January 3, 2024 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift settings of the required 12 S/RVs are within +/- 3% of the setpoint as follows:
In accordance with the INSERVICE TESTING PROGRAM Number of S/RVs 4
4 5
Setpoint (psig) 1135 1145 1155 Following testing, lift settings shall be within
+/- 1%.
NOTE------------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
Verify each required S/RV is capable of being opened in accordance with the INSERVICE TESTING PROGRAM.
OR Verify each required S/RV opens when manually actuated.
In accordance with the INSERVICE TESTING PROGRAM OPS as-left OPS pressures safety/relief valves (S/RVs)
+/- 1% of the nominal OPS Nominal Verify the as-found OPS lift pressures of the required S/RVs are within the limits specified in the COLR.
Reporting Requirements 5.6 (continued)
BFN-UNIT 1 5.0-24 Amendment No. 234, 239, 252, 281, 310, 325 January 13, 2023 5.6 Reporting Requirements (continued) 5.6.4 (Deleted).
5.6.5 CORE OPERATING LIMITS REPORT (COLR)
- a.
Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
(1)
The APLHGRs for Specification 3.2.1; (2)
The LHGR for Specification 3.2.3; (3)
The MINIMUM CRITICAL POWER RATIO (MCPR) and MCPR99.9% for Specification 3.2.2; (4)
The Manual Backup Stability Protection (BSP) Scram Region (Region I), the Manual BSP Controlled Entry Region (Region II), the modified APRM Flow Biased Simulated Thermal Power-High Scram setpoints used in the Automated BSP Scram Region, and the BSP Boundary for Specification 3.3.1.1; and (5)
The RBM setpoints and applicable reactor thermal power ranges for each of the setpoints for Specification 3.3.2.1, Table 3.3.2.1-1.
- b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1.
(Deleted).
- 2.
XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
(6) The as-found Overpressure Protection System lift pressures for Specification 3.4.3.
S/RVs 3.4.3 BFN-UNIT 2 3.4-7 Amendment 253 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety/Relief Valves (S/RVs)
LCO 3.4.3 The safety function of 12 S/RVs shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required S/RVs inoperable.
A.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND A.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OPS Overpressure Protection System (OPS)
OPS OPS inoperable.
S/RVs 3.4.3 BFN-UNIT 2 3.4-8 Amendment No. 255, 325, 338, 356 January 3, 2024 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift settings of the required 12 S/RVs are within +/- 3% of the setpoint as follows:
In accordance with the INSERVICE TESTING PROGRAM Number of S/RVs 4
4 5
Setpoint (psig) 1135 1145 1155 Following testing, lift settings shall be within
+/- 1%.
NOTE------------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
Verify each required S/RV is capable of being opened in accordance with the INSERVICE TESTING PROGRAM.
OR Verify each required S/RV opens when manually actuated.
In accordance with the INSERVICE TESTING PROGRAM OPS as-left OPS pressures safety/relief valves (S/RVs)
+/- 1% of the nominal OPS Nominal Verify the as-found OPS lift pressures of the required S/RVs are within the limits specified in the COLR.
Programs and Manuals 5.6 (continued)
BFN-UNIT 2 5.0-24 Amendment No. 287, 309, 333, 348 January 13, 2023 5.6 Reporting Requirements (continued) 5.6.4 (Deleted).
5.6.5 CORE OPERATING LIMITS REPORT (COLR) a.
Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
(1)
The APLHGRs for Specification 3.2.1; (2)
The LHGR for Specification 3.2.3; (3)
The MINIMUM CRITICAL POWER RATIO (MCPR) and MCPR 99.9% for Specification 3.2.2; (4)
The Manual Backup Stability Protection (BSP) Scram Region (Region I), the Manual BSP Controlled Entry Region (Region II), the modified APRM Flow Biased Simulated Thermal Power-High Scram setpoints used in the Automated BSP Scram Region, and the BSP Boundary for Specification 3.3.1.1; and (5)
The RBM setpoints and applicable reactor thermal power ranges for each of the setpoints for Specification 3.3.2.1, Table 3.3.2.1-1.
b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1.
XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
2.
XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986.
3.
EMF-85-74(P) Revision 0 Supplement 1(P)(A) and Supplement 2 (P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.
(6) The as-found Overpressure Protection System lift pressures for Specification 3.4.3.
S/RVs 3.4.3 BFN-UNIT 3 3.4-7 Amendment No. 212 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety/Relief Valves (S/RVs)
LCO 3.4.3 The safety function of 12 S/RVs shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required S/RVs inoperable.
A.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND A.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OPS Overpressure Protection System (OPS)
OPS OPS inoperable.
S/RVs 3.4.3 BFN-UNIT 3 3.4-8 Amendment No. 215, 285, 298, 316 January 3, 2024 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift settings of the required 12 S/RVs are within +/- 3% of the setpoint as follows:
In accordance with the INSERVICE TESTING PROGRAM Number of S/RVs 4
4 5
Setpoint (psig) 1135 1145 1155 Following testing, lift settings shall be within
+/- 1%.
NOTE------------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
Verify each required S/RV is capable of being opened in accordance with the INSERVICE TESTING PROGRAM.
OR Verify each required S/RV opens when manually actuated.
In accordance with the INSERVICE TESTING PROGRAM OPS as-left OPS pressures safety/relief valves (S/RVs)
+/- 1% of the nominal OPS Nominal Verify the as-found OPS lift pressures of the required S/RVs are within the limits specified in the COLR.
Reporting Requirements 5.6 (continued)
BFN-UNIT 3 5.0-24 Amendment No. 212, 226, 245, 250, 268, 293, 308 January 13, 2023 5.6 Reporting Requirements (continued) 5.6.4 (Deleted).
5.6.5 CORE OPERATING LIMITS REPORT (COLR) a.
Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
(1)
The APLHGRs for Specification 3.2.1; (2)
The LHGR for Specification 3.2.3; (3)
The MINIMUM CRITICAL POWER RATIO (MCPR) and MCPR99.9% for Specification 3.2.2; (4)
The Manual Backup Stability Protection (BSP) Scram Region (Region I), the Manual BSP Controlled Entry Region (Region II), the modified APRM Flow Biased Simulated Thermal Power-High Scram setpoints used in the Automated BSP Scram Region, and the BSP Boundary for Specification 3.3.1.1; and (5)
The RBM setpoints and applicable reactor thermal power ranges for each of the setpoints for Specification 3.3.2.1, Table 3.3.2.1-1.
b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1.
XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
2.
XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986.
3.
EMF-85-74(P) Revision 0 Supplement 1(P)(A) and Supplement 2 (P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.
(6) The as-found Overpressure Protection System lift pressures for Specification 3.4.3.
CNL-25-039 Revised Technical Specifications Bases Changes (Markups) for BFN Unit 1 (For Information Only)
(14 Pages)
RCS Pressure SL B 2.1.2 (continued)
BFN-UNIT 1 B 2.0-9 Revision 0 BASES (continued)
APPLICABLE The RCS safety/relief valves and the Reactor Protection SAFETY ANALYSES System Reactor Vessel Steam Dome Pressure - High Function have settings established to ensure that the RCS pressure SL will not be exceeded.
The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME, Boiler and Pressure Vessel Code, 1965 Edition, including Addenda through the summer of 1965 (Ref. 5), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig. The SL of 1325 psig, as measured in the reactor steam dome is equivalent to 1375 psig at the lowest elevation of the RCS. The RCS is designed to the USAS Nuclear Power Piping Code, Section B31.1, 1967 Edition (Ref. 6), and the additional requirements of GE design and procurement specifications (Ref. 7) which were implemented in lieu of the outdated B31 Nuclear Code Cases - N2, N7, N9, and N10, for the reactor recirculation piping, which permits a maximum pressure transient of 120% of design pressures of 1148 psig for suction piping and 1326 psig for discharge piping. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes.
SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings is 120% of design pressures of 1148 psig for suction piping and 1326 psig for discharge piping.
Overpressure Protection System
Control Rod Scram Times B 3.1.4 (continued)
BFN-UNIT 1 B 3.1-27 Revision 0, 68 October 18, 2012 BASES (continued)
APPLICABLE The analytical methods and assumptions used in evaluating the SAFETY ANALYSES control rod scram function are presented in References 2, 3, and 4. The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. The resulting negative scram reactivity forms the basis for the determination of plant thermal limits (e.g., the MCPR). Other distributions of scram times (e.g., several control rods scramming slower than the average time with several control rods scramming faster than the average time) can also provide sufficient scram reactivity. Surveillance of each individual control rod's scram time ensures the scram reactivity assumed in the DBA and transient analyses can be met.
The scram function of the CRD System protects the MCPR Safety Limit (SL) (see Bases for SL 2.1.1, "Reactor Core SLs,"
and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)") and the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)", and LCO 3.2.3, LINEAR HEAT GENERATION RATE (LHGR)), which ensure that no fuel damage will occur if these limits are not exceeded. Above 800 psig, the scram function is designed to insert negative reactivity at a rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL, during the analyzed limiting power transient. Below 800 psig, the scram function is assumed to perform during the control rod drop accident (Refs. 5, 8, and 9) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO 3.1.6, "Rod Pattern Control").
For the reactor vessel overpressure protection analysis, the scram function, along with the safety/relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.
Control rod scram times satisfy Criterion 3 of the NRC Policy Statement (Ref. 7).
Overpressure Protection System
RPS Instrumentation B 3.3.1.1 (continued)
BFN-UNIT 1 B 3.3-14 Revision 0, 40 October 26, 2006 BASES APPLICABLE 2.c. Average Power Range Monitor Fixed Neutron Flux - High SAFETY ANALYSES, LCO, and The Average Power Range Monitor Fixed Neutron Flux - High APPLICABILITY Function is capable of generating a trip signal to prevent fuel damage or excessive RCS pressure. For the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Fixed Neutron Flux - High Function is assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the safety/relief valves (S/RVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 5) takes credit for the Average Power Range Monitor Fixed Neutron Flux - High Function to terminate the CRDA.
The Allowable Value is based on the Analytical Limit assumed in the CRDA analyses.
The Average Power Range Monitor Fixed Neutron Flux - High Function is required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being exceeded.
Although the Average Power Range Monitor Fixed Neutron Flux - High Function is assumed in the CRDA analysis, which is applicable in MODE 2, the Average Power Range Monitor Neutron Flux - High, (Setdown) Function conservatively bounds the assumed trip and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average Power Range Monitor Fixed Neutron Flux - High Function is not required in MODE 2.
Overpressure Protection System
RPS Instrumentation B 3.3.1.1 (continued)
BFN-UNIT 1 B 3.3-17 Revision 0 BASES APPLICABLE
- 3. Reactor Vessel Steam Dome Pressure - High SAFETY ANALYSES, (PIS-3-22AA, PIS-3-22BB, PIS-3-22C, and PIS-3-22D)
LCO, and APPLICABILITY An increase in the RPV pressure during reactor operation (continued) compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. The Reactor Vessel Steam Dome Pressure - High Function initiates a scram for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection analysis of Reference 4, reactor scram (the analyses conservatively assume scram on the Average Power Range Monitor Fixed Neutron Flux - High signal, not the Reactor Vessel Steam Dome Pressure - High signal), along with the S/RVs, limits the peak RPV pressure to less than the ASME Section III Code limits.
High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure. The Reactor Vessel Steam Dome Pressure - High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event.
Four channels of Reactor Vessel Steam Dome Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required to be OPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists.
Overpressure Protection System
RPS Instrumentation B 3.3.1.1 (continued)
BFN-UNIT 1 B 3.3-19a Revision 41 November 09, 2006 BASES APPLICABLE
- 4. Reactor Vessel Water Level - Low, Level 3 SAFETY ANALYSES, (LIS-3-203A, LIS-3-203B, LIS-3-203C, and LIS-3-203D)
LCO, and (continued)
APPLICABILITY Corrective Action Program, additional evaluations and potential corrective actions will be performed as necessary to ensure that any as-found setting, which is conservative to the Allowable Value, but outside the acceptable as-found band is evaluated for long-term reliability trends.
- 5. Main Steam Isolation Valve - Closure MSIV closure results in loss of the main turbine and the condenser as a heat sink for the nuclear steam supply system and indicates a need to shut down the reactor to reduce heat generation. Therefore, a reactor scram is initiated on a Main Steam Isolation Valve - Closure signal before the MSIVs are completely closed in anticipation of the complete loss of the normal heat sink and subsequent overpressurization transient.
However, for the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Fixed Neutron Flux - High Function, along with the S/RVs, limits the peak RPV pressure to less than the ASME Code limits. That is, the direct scram on position switches for MSIV closure events is not assumed in the overpressurization analysis. Additionally, MSIV closure is assumed in the transients analyzed in Reference 7 (e.g., low steam line pressure, manual closure of MSIVs, high steam line flow).
The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
Overpressure Protection System
ATWS-RPT Instrumentation B 3.3.4.2 (continued)
BFN-UNIT 1 B 3.3-121 Revision 0 BASES APPLICABLE
- b. Reactor Steam Dome Pressure - High SAFETY ANALYSES, (PIS-3-204A, PIS-3-204B, PIS-3-204C, and PIS-3-204D)
LCO, and APPLICABILITY Excessively high RPV pressure may rupture the RCPB. An (continued) increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization. The Reactor Steam Dome Pressure - High Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation. For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the safety/relief valves, limits the peak RPV pressure to less than the ASME Section III Code limits.
The Reactor Steam Dome Pressure - High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure. Four channels of Reactor Steam Dome Pressure -
High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Steam Dome Pressure - High Allowable Value is chosen to provide an adequate margin to the ASME Section III Code limits.
Overpressure Protection System
S/RVs B 3.4.3 (continued)
BFN-UNIT 1 B 3.4-17 Revision 0 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.3 Safety/Relief Valves (S/RVs)
BASES BACKGROUND The ASME Boiler and Pressure Vessel Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of S/RVs are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).
The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell.
The S/RVs can actuate by either of two modes: the safety mode or the relief mode. In the safety mode (or spring mode of operation), the spring loaded pilot valve opens when steam pressure at the valve inlet overcomes the spring force holding the pilot valve closed. Opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve. This satisfies the Code requirement.
Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool. The S/RVs that provide the relief mode are the Automatic Depressurization System (ADS) valves. ADS requirements are specified in LCO 3.5.1, "ECCS - Operating."
OPS Overpressure Protection System (OPS)
American Society of Mechanical Engineers (ASME)
(Ref. 2) safety/relief valves (S/RVs)
The Overpressure Protection System (OPS) prevents overpressurization of the nuclear system by discharging reactor steam to the suppression pool. This action protects the reactor coolant pressure boundary (RCPB) from failure which could result in the release of fission products (Ref. 1).
Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool.
The safety mode is credited for overpressure protection.
In the relief mode (or power actuated mode of operation), a pneumatic piston or cylinder and mechanical linkage assembly are used to open the valve by overcoming the spring force at valve inlet pressure above 50 psig. The pneumatic operator is arranged so that its malfunction will not prevent the valve disk from lifting if steam inlet pressure reaches the spring lift set pressures. In the relief mode, valves may be opened manually or automatically at the selected preset pressure. S/RVs operating in relief mode are not credited for overpressure protection.
Some of the S/RVs operating in relief mode also provide the Automatic Depressurization System (ADS) function, specified in LCO 3.5.1, "ECCS - Operating." The instrumentation associated with the relief mode and the ADS function is discussed in LCO 3.3.5.1, "Emergency Core Cooling Systems (ECCS) Instrumentation."
The
S/RVs B 3.4.3 (continued)
BFN-UNIT 1 B 3.4-18 Revision 0 BASES (continued)
APPLICABLE The overpressure protection system must accommodate the SAFETY ANALYSES most severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref. 1). For the purpose of the analyses, 12 S/RVs are assumed to operate in the safety mode. The analysis results demonstrate that the design S/RV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig =
1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event.
Reference 2 discusses additional events that are expected to actuate the S/RVs. From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above.
S/RVs satisfy Criterion 3 of the NRC Policy Statement (Ref. 4).
LCO The safety function of 12 S/RVs are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs. 1 and 2).
The requirements of this LCO are applicable only to the capability of the S/RVs to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety function).
The S/RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the OPS OPS The S/RV discharge piping is designed to accommodate forces resulting from the relief action including interactions with the suppression pool and is supported for reactions due to flow at maximum S/RV discharge capacity. The overpressure protection (Ref. 1) assume twelve of operation OPS design basis event 3
The OPS satisfies to be set The OPS is OPERABLE when it can ensure that the ASME Code limit on peak reactor pressure, as stated in Safety Limit 2.1.2, will be protected using the safety and relief modes mode of the S/RVs. The OPERABILITY of the OPS is only dependent on the ability to relieve excess pressure and maintain reactor pressure below Safety Limit 2.1.2, and may credit less than the full complement of installed S/RVs.
S/RVs B 3.4.3 (continued)
BFN-UNIT 1 B 3.4-19 Revision 0 BASES LCO highest safety valve to be set so that the total accumulated (continued) pressure does not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the FSAR are based on these setpoints, but also include the additional uncertainties of +/- 3% of the nominal setpoint drift to provide an added degree of conservatism.
Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded.
APPLICABILITY In MODES 1, 2, and 3, all required S/RVs must be OPERABLE, since considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat.
In MODE 4, decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents. In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The S/RV function is not needed during these conditions.
overpressure Reference 3 An inoperable OPS Safety Limit 2.1.2 the OPS there may be OPS OPS OPS
S/RVs B 3.4.3 (continued)
BFN-UNIT 1 B 3.4-20 Revision 0, 109 November 30, 2017 BASES (continued)
ACTIONS A.1 and A.2 With less than the minimum number of required S/RVs OPERABLE, a transient may result in the violation of the ASME Code limit on reactor pressure. If the safety function of one or more required S/RVs is inoperable, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.4.3.1 REQUIREMENTS This Surveillance requires that the 12 required S/RVs open at the pressures assumed in the safety analysis of Reference 1.
The setpoint groups for all 13 S/RVs are listed. The demonstration of the S/RV safe lift settings must be performed during shutdown, since this is a bench test, to be done in accordance with the INSERVICE TESTING PROGRAM. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RV setpoint tolerance is +/- 3% for OPERABILITY; however, the valves are reset to +/- 1% during the Surveillance to allow for drift.
If the OPS is inoperable This Surveillance verifies that the S/RVs credited by the OPS have their as-left settings within 1% of the nominal opening pressure setpoint. The OPS may credit less than the full complement of installed S/RVs and the SR only applies to those S/RVs required to meet the LCO. The verification of the S/RV as-left settings is performed in accordance with the INSERVICE TESTING PROGRAM. The nominal setpoint shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RVs are reset to +/- 1% during the Surveillance to allow for drift.
S/RVs B 3.4.3 (continued)
BFN-UNIT 1 B 3.4-21 Revision 0 Amendment 333 January 3, 2024 BASES SURVEILLANCE SR 3.4.3.2 REQUIREMENTS (continued)
This Surveillance verifies that each S/RV is capable of being opened, which can be determined by either of the following two methods.
Method 1 Valve OPERABILITY and setpoints for overpressure protection are verified in accordance with the requirements of the ASME OM Code (Ref. 3). Proper S/RV function is verified through performance of inspections and overlapping tests on component assemblies, demonstrating the valve is capable of being opened. Testing is performed to demonstrate that:
Each S/RV main stage opens and passes steam when the associated pilot stage actuates;
Each S/RV pilot stage actuates to open the associated main stage when the pneumatic actuator is pressurized;
Each S/RV solenoid valves ports pneumatic pressure to the associated S/RV actuator when energized; and
Each S/RV actuator stem moves when dry lift tested in-situ.
With exception of the main and pilot stages this test demonstrates mechanical operation without steam.
The solenoid valves and S/RV actuators are functionally tested as part of the INSERVICE TESTING PROGRAM. The S/RV assembly is bench tested as part of the certification process, at intervals determined in accordance with the INSERVICE TESTING PROGRAM. Maintenance procedures ensure that the S/RV is correctly installed in the plant and that the S/RV and associated piping remain clear of foreign material that might obstruct valve operation or full steam flow.
This Surveillance verifies that the as-found lift pressures of the S/RVs credited by the OPS are consistent with the assumptions of the overpressure analysis. The OPS may credit less than the full complement of installed S/RVs required to meet the LCO. Any valve not credited for SR 3.4.3.1 may not be credited for this Surveillance in the same operating cycle. The measurement of the S/RV lift pressures must be performed in accordance with the INSERVICE TESTING PROGRAM. The OPS S/RV lift pressures are specified in the COLR.
S/RVs B 3.4.3 (continued)
BFN-UNIT 1 B 3.4-21a Revision 0 Amendment 333 January 3, 2024 BASES SURVEILLANCE SR 3.4.3.2 (continued)
REQUIREMENTS (continued)
Method 2 A manual actuation of each required S/RV is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or by any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the S/RVs divert steam flow upon opening. Sufficient time is therefore allowed after the required pressure is achieved to perform this test. Adequate pressure at which this test is to be performed is 920 psig (the pressure recommended by the valve manufacturer). Adequate steam flow is represented by at least 3 turbine bypass valves open. Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME Code requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR. If a valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the S/RV is considered OPERABLE.
S/RVs B 3.4.3 BFN-UNIT 1 B 3.4-22 Revision 0, 43, 81, 123 Amendment 333 January 3, 2024 BASES SURVEILLANCE SR 3.4.3.2 (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the INSERVICE TESTING PROGRAM. Industry operating experience has shown that these components usually pass the SR when performed at the Code required Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
REFERENCES
- 1. FSAR, Section 4.4.6.
- 2. FSAR, Section 14.5.1.
- 4. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
3
- 2. ASME Boiler and Pressure Vessel Code,Section III.
Reactor Steam Dome Pressure B 3.4.10 (continued)
BFN-UNIT 1 B 3.4-67 Revision 0, 50 May 03, 2007 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.10 Reactor Steam Dome Pressure BASES BACKGROUND The reactor steam dome pressure is an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria and is also an assumed initial condition of design basis accidents and transients.
APPLICABLE The reactor steam dome pressure of 1055 psig is an initial SAFETY ANALYSES condition of the vessel overpressure protection analysis of Reference 1. This analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the pressure relief system, primarily the safety/relief valves, during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved. Reference 2 also assumes an initial reactor steam dome pressure for the analysis of design basis accidents and transients used to determine the limits for fuel cladding integrity (see Bases for LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)") and 1% cladding plastic strain (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"). Since the design basis accident and the transient analyses are performed at nominal operating pressures (1035 psig), a Overpressure Protection System CNL-25-039 Example Updated Core Operating Limits Report for BFN Unit 1 (For Information Only)
(1 Page)
IIVA Reactor Engineering and Fuels - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: Month DD, YYYY Browns Ferry Unit 1 Cycle 16 Core Operating Limits Report, (120% OLTP, MELLLA+)
Page XX TVA-COLR-BF1C16, Revision 1 XX Overpressure Protection System (Technical Specification 3.4.3)
The as-found OPS lift pressures of the required safety/relief valves are specified below:
Number of S/RVs 3
4 5
As-Found Lift Pressure Limit (psig) 1169 1179 1189 Proposed Revision to BFN Unit 1 COLR