CNL-24-049, Brows Ferry Nuclear Plant, Units 1, 2 and 3 - License Amendment Request for Adoption of TSTF-576 to Revise Safety/Relief Valve Requirements
| ML24344A034 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  (DPR-033, DPR-052, DPR-068) |
| Issue date: | 12/09/2024 |
| From: | Hulvey K Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| CNL-24-049 | |
| Download: ML24344A034 (1) | |
Text
1101 Market Street, Chattanooga, Tennessee 37402 CNL-24-049 December 9, 2024 10 CFR 50.90 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 - License Amendment Request for Adoption of TSTF-576 to Revise Safety/Relief Valve Requirements In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, Application for amendment of license, construction permit, or early site permit, Tennessee Valley Authority (TVA) is submitting a request for an amendment to Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 for the Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, respectively.
TVA requests adoption of Technical Specifications Task Force (TSTF)-576-A, Revision 3, Revise Safety/Relief Valve Requirements. The proposed change revises the safety/relief valve (S/RV)
TS to align the overpressure protection requirements with the safety limits and regulations.
The enclosure provides a description and assessment of the proposed change. Attachment 1 provides the existing TS pages marked-up to show the proposed changes. Attachment 2 provides the existing BFN Unit 1 TS Bases pages marked-up to show the proposed changes (BFN Units 2 and 3 changes will be similar). Changes to the existing TS Bases are provided for information only and will be implemented under the Technical Specification Bases Control Program. Attachment 3 provides an example revised BFN Unit 1 Core Operating Limits Report, for information only, that illustrates the addition of the S/RV limits (BFN Units 2 and 3 changes will be similar).
TVA requests that the amendment be reviewed under the Consolidated Line Item Improvement Process. Approval of the proposed amendment is requested within 6 months of acceptance.
Once approved, the amendment shall be implemented within 90 days.
In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Alabama official.
U. S. Nuclear Regulatory Commission CNL-24-049 Page 2 December 9, 2024 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 9th day of December 2024.
Respectfully, Kimberly D. Hulvey General Manager, Nuclear Regulatory Affairs & Emergency Preparedness
Enclosure:
Description and Assessment of the Proposed Change 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: 2024.12.09 05:45:03
-05'00'
Enclosure CNL-24-049 E1 of 6 Description and Assessment of the Proposed Change
Subject:
Browns Ferry Nuclear Plant, Units 1, 2, and 3 - License Amendment Request for Adoption of TSTF-576 to Revise Safety/Relief Valve Requirements CONTENTS
1.0 DESCRIPTION
............................................................................................................ 2
2.0 ASSESSMENT
............................................................................................................ 2 2.1 Applicability of Safety Evaluation............................................................................ 2 2.2 Optional Changes and Variations............................................................................ 2
3.0 REGULATORY ANALYSIS
......................................................................................... 4 3.1 No Significant Hazards Consideration Analysis...................................................... 4 3.2 Conclusion............................................................................................................... 5 4.0 ENVIRONMENTAL EVALUATION............................................................................. 6 Attachments:
- 1. Proposed TS Changes (Markups) for BFN Units 1, 2, and 3
- 2. Proposed TS Bases Changes (Markups) for BFN Unit 1 (For Information Only)
- 3. Example Updated Core Operating Limits Report for BFN Unit 1 (For Information Only)
Enclosure CNL-24-049 E2 of 6 Description and Assessment of the Proposed Change
1.0 DESCRIPTION
Tennessee Valley Authority (TVA) requests adoption of Technical Specifications Task Force (TSTF)-576-A, Revision 3, Revise Safety/Relief Valve Requirements. The proposed change revises the safety/relief valve (S/RV) technical specifications (TS) to align the overpressure protection requirements with the safety limits and the regulations.
2.0 ASSESSMENT
2.1 Applicability of Safety Evaluation TVA has reviewed the safety evaluation for TSTF-576-A, Revision 3, provided to the Technical Specifications Task Force in a letter dated September 10, 2024. This review included a review of the Nuclear Regulatory Commission (NRC) staffs evaluation, as well as the information provided in TSTF-576-A, Revision 3. As described herein, TVA has concluded that the justifications presented in TSTF-576-A, Revision 3, and the safety evaluation prepared by the NRC staff are applicable to Browns Ferry Nuclear Plant (BFN),
Units 1, 2, and 3 and justify this amendment for the incorporation of the changes to the BFN TS.
The NRC-approved overpressure protection analysis methodology for BFN is Framatome Inc. topical report ANP-10300P-A, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Transient and Accident Scenarios.
2.2 Optional Changes and Variations TVA is proposing the following variations from the TS changes described in TSTF-576-A, Revision 3, or the applicable parts of the NRC staffs safety evaluation:
TSTF-576 revises the Standard Technical Specifications (STS) table of contents (TOC) to change STS 3.4.3, Safety/Relief Valves to Overpressure Protection System. The TOC of the BFN TS is licensee-controlled and is not included in Attachments 1 and 2.
TSTF-576 revises Condition A, B, and C of STS 3.4.3, Safety/Relief Valves. The BFN TS contain requirements for safety/relief valves (S/RVs) that differ from the STS on which TSTF-576 was based but are encompassed in the TSTF-576 justification.
Condition C that is revised in STS 3.4.3 is analogous to Condition A in the BFN TS.
The BFN TS does not contain any Conditions corresponding to Condition A or B of STS 3.4.3.
TSTF-576 revises Surveillance Requirement (SR) 3.4.3.1 by deleting a Note that is not contained in, and never was part of, the BFN TS. The remaining changes to SR 3.4.3.1 are applicable to the BFN TS.
TSTF-576 revises SR 3.4.3.2 by replacing the surveillance, in its entirety, with new language that directs operators to the Core Operating Limits Report (COLR) for as-found lift pressure limits. BFN SR 3.4.3.2 currently contains an additional OR
Enclosure CNL-24-049 E3 of 6 statement which states Verify each required S/RV is capable of being opened in accordance with the INSERVICE TESTING PROGRAM, which was added by Amendments 333, 356, and 316 (ML23319A199). This additional SR is no longer needed based on the TSTF. Accordingly, this SR and its associated Bases are being deleted. The term required S/RV is retained from the current BFN TS in accordance with TSTF-576. These changes differ from the STS on which TSTF-576 was based but are encompassed in the TSTF-576 justification.
TSTF-576 revises Bases 3.3.6.3, Low-Low Set (LLS) Instrumentation. The BFN TS does not have an analogous TS for LLS valves, and this change cannot be implemented.
TSTF-576 revises Bases 3.4.3 as follows:
o The bracketed language in the Bases background regarding S/RVs operating in relief mode being credited for overpressure protection is not applicable to BFN because only the safety mode (self-actuation) is utilized for overpressure protection. Due to BFN S/RV design, the pressure needed for relief mode to overcome spring force is 50 psig instead of the TSTF pressure of 0 psig. Regarding the bracketed paragraph being added to the end of the background section, the BFN TS does not contain LLS valves. However, the Bases language regarding the Automatic Depressurization System is applicable to BFN and is adopted from TSTF-576. The TSTF traveler erroneously references the American Society of Mechanical Engineers (ASME) Operations and Maintenance Code (Ref. 2) for the ASME Boiler and Pressure Vessel Code (BPVC). Reference 2 is being changed to ASME BPVC,Section III, as that contains the correct design requirements for the BFN reactor pressure vessel.
o The STS value of 11 S/RVs is bracketed in the revision to the applicable safety analyses section of Bases 3.4.3. The corresponding value for BFN is 12 S/RVs. The reference to NRC No.93-102 in the applicable safety analyses section has been replaced by 10 CFR 50.36(c)(2)(ii) to match the STS, and Reference 4 of Bases 3.4.3 is deleted accordingly.
o 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.
o The revised Bases Actions A.1 and A.2 delete [three] or more [required]
S/RVs are inoperable, which corresponds to the BFN TS Bases Actions C.1 and C.2 deletion of with less than the minimum number of required S/RVs OPERABLE. These changes differ from the STS on which TSTF-576 was based but are encompassed in the TSTF-576 justification.
TSTF-576 revises Bases 3.4.11, Reactor Steam Dome Pressure. However, the analogous section in the BFN TS Bases is 3.4.10. This difference is administrative and is a non-technical variation to TSTF-576.
Enclosure CNL-24-049 E4 of 6
3.0 REGULATORY ANALYSIS
3.1 No Significant Hazards Consideration Analysis Tennessee Valley Authority (TVA) requests adoption of Technical Specifications Task Force (TSTF)-576-A, Revision 3, Revise Safety/Relief Valve Requirements. The proposed change revises the safety/relief valve (S/RV) technical specifications (TS) to align the overpressure protection requirements with the safety limits and the regulations. The limiting condition for operation (LCO) is revised to replace requirements on each credited S/RV with a requirement that the overpressure protection system (OPS) be operable. The surveillance requirements (SRs) are revised to move the as-found S/RV lift pressure limits to the licensee-controlled Core Operating Limits Report (COLR). An SR that tests the ability of the S/RVs to be capable of manual operation is removed as that capability is not credited in any safety analysis. The TS Actions are revised to be consistent with the changes to the LCO and SRs. Administrative changes are made to the TS for clarity and consistency. The COLR specification is revised to reference the OPS specification.
TVA has evaluated if a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in Title 10 of the Code of Federal Regulations (10 CFR) 50.92, Issuance of amendment, as discussed below:
- 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed change revises the S/RV TS to align the overpressure protection requirements with the safety limits and the regulations. The overpressure protection system must accommodate the 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). The proposed change does not affect the MSIVs and would have no effect on the probability of the MSIVs closing or generation of a reactor scram signal on MSIV position. Therefore, the probability of the event is unaffected. The consequences of the accident are based on the peak reactor pressure vessel pressure. Both the current and proposed TS ensure the overpressure safety limit is not exceeded. The accident analyses consider the aggregate operation of the credited S/RVs, not the performance of individual valves. The proposed change moves the S/RV as-found lift pressure limits to the licensee controlled COLR which uses Nuclear Regulatory Commission (NRC)-approved methodologies.
Altering the control process for these values has no effect on the accident evaluations. As a result, the consequences of the accident are not changed.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No
Enclosure CNL-24-049 E5 of 6 The proposed change revises the S/RV TS to align the overpressure protection requirements with the safety limits and the regulations. The proposed change does not alter the design function or operation of the S/RVs. The proposed change does not create any new credible failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing basis.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No The proposed change revises the S/RV TS to align the overpressure protection requirements with the safety limits and the regulations. The proposed change ensures that the S/RVs can protect Safety Limit 2.1.2. Although the as-found S/RV lift pressure limits are moved to the licensee-controlled COLR, the safety margin provided by the S/RVs, which ensures the safety limit is protected, is not changed. The conservatisms in the evaluation and the analysis are described in the NRC-approved methods for each licensee, which are not altered by the proposed change. The proposed change does not alter a design basis limit or a safety limit, and, therefore, does not reduce the margin of safety.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, TVA concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.
3.2 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) 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.
Enclosure CNL-24-049 E6 of 6 4.0 ENVIRONMENTAL EVALUATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.
However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
Enclosure CNL-24-049 Proposed Technical Specifications Changes (Markups) for BFN Units 1, 2, and 3 (6 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 limits specified in the COLR.
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 limits specified in the COLR.
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 limits specified in the COLR.
Enclosure CNL-24-049 Proposed 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 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.
10 CFR 50.36(c)(2)(ii)
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. 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
Enclosure CNL-24-049 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)
For INSERVICE TESTING PROGRAM requirements, the as-found set-pressure acceptance criteria for safety/relief valves (S/RVs) is +/- 3% of valve nameplate set-pressure.
The as-found OPS lift pressures of the required S/RVs are within +/- 3% of the limits specified below:
Number of S/RVs 4
4 5
As-Found Lift Pressure Limit (psig) 1135 1145 1155 Proposed Revision to BFN Unit 1 COLR