CNL-23-018, – Request for Amendment Regarding Technical Specification Surveillance Requirements 3.4.3.2 and 3.5.1.11 Regarding Safety Relief Valves (BFN TS-540)
| ML23089A167 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/30/2023 |
| From: | Hulvey K Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| CNL-23-018 | |
| Download: ML23089A167 (1) | |
Text
1101 Market Street, Chattanooga, Tennessee 37402 CNL-23-018 March 30, 2023 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 - Request for Amendment Regarding Technical Specification Surveillance Requirements 3.4.3.2 and 3.5.1.11 Regarding Safety Relief Valves (BFN TS-540)
In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, Tennessee Valley Authority (TVA) is submitting for Nuclear Regulatory Commission (NRC) approval, 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. TVA requests a change to the BFN Units 1, 2, and 3 Technical Specification (TS)
Surveillance Requirements (SRs) 3.4.3.2 and 3.5.1.11 regarding safety relief valves (SRV). The change would supplement the current requirement in these TS SRs to verify the safety relief valves open when manually actuated with an alternate requirement that verifies that the SRVs are capable of being opened and also revises the SR frequency to be in accordance with the Inservice Testing Program.
The enclosure provides a description and evaluation of the proposed change. Attachment 1 to the enclosure provides the existing TS pages for BFN Units 1, 2, and 3, respectively, marked to show the proposed change. Attachment 2 to the enclosure provides the proposed Units 1, 2, and 3 TS pages retyped to show the changes incorporated. Attachment 3 to the enclosure provides TS Bases pages for BFN Unit 1 (the BFN Units 2 and 3 Bases are nearly identical).
Changes to the existing TS Bases are provided for information only and will be implemented under the TS Bases Control Program.
TVA has determined that there are no significant hazards considerations associated with the proposed change and that the TS change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). In accordance with 10 CFR 50.91(b)(1), TVA is providing a copy of the proposed license amendment to the Alabama Department of Public Health.
U.S. Nuclear Regulatory Commission CNL-23-018 Page 2 March 30, 2023 Due to limited resources related to refueling outage activities, TVA was unable to submit this request in time to provide a one-year review period. The generation forecast for BFN was also adjusted during development of this request, which moved the start date for the Unit 3 Cycle 21 refueling outage to earlier than was previously planned. Therefore, TVA requests an expedited NRC approval of this request by February 15, 2024, with implementation to be completed prior to the completion of the Unit 3 Cycle 21 refueling outage in spring 2024.
There are no new regulatory commitments associated with this submittal. Please address any questions regarding this request to Stuart L. Rymer, Senior Manager, Fleet Licensing, at slrymer@tva.gov.
I declare under penalty of perjury that the foregoing is true and correct. Executed on this 30th day of March 2023.
Respectfully, Kimberly D. Hulvey Director, Nuclear Regulatory Affairs
Enclosure:
Description and Assessment of the Proposed Changes cc (with 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: 2023.03.30 09:35:59 -04'00'
Enclosure Description and Assessment of the Proposed Changes CNL-23-018 E1 of 10
Subject:
Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Request for Amendment Regarding Technical Specification Surveillance Requirements 3.4.3.2 and 3.5.1.11 Regarding Safety Relief Valves (BFN TS-540)
CONTENTS 1.0
SUMMARY
DESCRIPTION.............................................................................................. 2 2.0 DETAILED DESCRIPTION............................................................................................... 2 2.1 System Design and Operation....................................................................................... 2 2.2 Current Technical Specifications Requirements............................................................ 2 2.3 Reason for the Proposed Change................................................................................. 3 2.4 Description of the Proposed Change............................................................................. 3
3.0 TECHNICAL EVALUATION
.............................................................................................. 4
4.0 REGULATORY EVALUATION
.......................................................................................... 6 4.1 Applicable Regulatory Requirements............................................................................ 6 4.2 Precedent...................................................................................................................... 7 4.3 No Significant Hazards Consideration Determination.................................................... 7 4.4 Conclusions................................................................................................................... 9
5.0 ENVIRONMENTAL CONSIDERATION
............................................................................ 9
6.0 REFERENCES
................................................................................................................ 10 Attachments:
- 1.
Proposed Technical Specification Changes (Markups) for BFN Units 1, 2, and 3
- 2.
Proposed Technical Specification Changes (Final Typed) for BFN Units 1, 2, and 3
- 3.
Proposed Technical Specification Bases Changes (Markups) for BFN Unit 1 (For Information Only)
Enclosure CNL-23-018 E2 of 10 1.0
SUMMARY
DESCRIPTION Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.90, 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. The proposed change revises the BFN Units 1, 2, and 3 Technical Specifications (TS) 3.4.3, Safety Relief Valves (SRVs), Surveillance Requirement (SR) 3.4.3.2 and BFN Units 1, 2, and 3 TS 3.5.1, ECCS - Operating, SR 3.5.1.11 by replacing the current requirement to verify the SRVs open when manually actuated with an alternate requirement that verifies the SRVs are capable of being opened. The proposed change also revises the frequency for these SRs to be in accordance with the Inservice Testing (IST) Program.
2.0 DETAILED DESCRIPTION 2.1 System Design and Operation The nuclear system pressure relief system includes 13 SRVs each for Units 1, 2, and 3, which are located on the main steam lines within the drywell between the reactor vessel and the first isolation valve. The SRVs are Target Rock Model 7567F two-stage, pilot-operated safety/relief valves. They are Code Class 1 valves categorized in accordance with ISTC-1300 as Category B/C because they are capable of remote-manual operation when inlet pressure is below valve setpoint and are self-actuated when inlet pressure reaches valve setpoint. They are mounted on the four main steam lines so that a single accident cannot completely disable a safety, relief, or automatic depressurization function. The SRVs are installed so that each valve discharge is piped through its own discharge line to a point below the minimum water level in the primary containment suppression pool to permit the steam to condense in the pool.
The SRVs can actuate by either of two modes: the safety mode or the remote actuation 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. Remote actuation mode energizes the solenoid valve and ports air to the pilot valve, opening the pilot valve which opens the main valve. The valve can be remotely actuated by a manual hand switch or through automatic depressurization system (ADS) logic.
Six of the SRVs serve as ADS valves. Their purpose is to automatically depressurize the reactor vessel thus allowing injection by low pressure emergency cooling sources.
2.2 Current Technical Specifications Requirements SR 3.4.3.2
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 is adequate to perform the test.
Verify each required SRV opens when manually actuated.
Enclosure CNL-23-018 E3 of 10 SR 3.5.1.11
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 is adequate to perform the test.
Verify each required ADS valve opens when manually actuated.
The current frequency for each SR is in accordance with the Surveillance Frequency Control Program (SFCP).
2.3 Reason for the Proposed Change TS surveillance testing of SRVs at BFN is currently performed at a reactor pressure vessel (RPV) pressure greater than or equal to 920 pounds per square inch gauge (psig) to verify that mechanically, each valve is functioning properly and no blockage exists in the valve discharge line.
The Boiling Water Reactor Owners' Group (BWROG) Evaluation of NUREG-0737, Clarification of TMI Action Plan Requirements, Item II.K.3.16, Reduction of Challenges and Failures of Relief Valves - Feasibility Study and System Modification, (Reference 1) recommends avoiding unnecessary challenges to the SRVs and modifying operations to reduce spurious SRV blowdowns and to reduce the probability of SRVs to stick open when challenged.
Experience in the industry has shown that manual actuation of SRVs during plant operation may create a potential for SRV seat leakage. Potential SRV leakage is routed to the suppression pool. The increased heat and fluid additions to the suppression pool requires more frequent suppression pool cooling and more frequent pump-down operations to control suppression pool temperature and level. Main stage SRV seat leakage also tends to mask the indications of SRV pilot stage seat-leakage. Pilot stage leakage could cause spurious SRV actuation and/or SRV failure to reclose after actuation. Excessive leakage would require plant shutdown to replace the leaking SRV.
Reducing or eliminating the number of manual actuations of the SRVs during plant startup minimizes the potential depressurization and cooldown events due to failure-to-close SRV events and minimizes the potential for pilot or main stage leakage of the SRVs. Implementing this change would still maintain the capability to manually open and close SRVs, as necessary, for the IST program or as corrective action for SRVs with excessive leakage.
2.4 Description of the Proposed Change The proposed change revises BFN Units 1, 2, and 3 SR 3.4.3.2 to verify each required SRV is capable of being opened as an option to verifying that each required SRV opens. Similarly, BFN Units 1, 2, and 3 SR 3.5.1.11 is revised to verify each required ADS valve is capable of being opened as an option to verifying that each required ADS valve opens. Additionally, the proposed change revises the frequency for performing these SRs to be in accordance with the IST Program. The proposed TS changes are similar to those previously approved by the NRC in References 2, 3, and 4. to this enclosure provides the existing TS pages for BFN Units 1, 2, and 3, respectively, marked to show the proposed change. Attachment 2 to this enclosure provides the proposed Units 1, 2, and 3 TS pages retyped to show the changes incorporated.
Enclosure CNL-23-018 E4 of 10 to this enclosure provides TS Bases pages for BFN Unit 1 (the BFN Units 2 and 3 Bases are nearly identical). Changes to the existing TS Bases are provided for information only and will be implemented under the TS Bases Control Program.
3.0 TECHNICAL EVALUATION
The manual actuation tests currently prescribed in SRs 3.4.3.2 and 3.5.1.11 provide demonstration of the mechanical operation of the SRVs and overlaps with other testing to demonstrate that the functions of the SRVs can be performed. These manual actuation tests are currently performed once per operating cycle (24 months) corresponding to start-up from refueling outages. The SRV manual actuation lift test is credited for demonstrating the mechanical functioning of the valve for the remote actuation mode and for the automatic depressurization function.
BFN Units 1, 2, and 3 are in the fourth ten-year IST interval, and the Code of Record (COR) is the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), 2004 Edition through 2006 Addenda1. Mandatory Appendix I, paragraph l-3410(c), of the COR permits the valve disk stroke capability to be verified by mechanical examination or tests. Overlapping tests can be credited to individually test SRV components. The proposed amendment revises the TS to reflect the testing provisions of the COR, while retaining the current TS capability to manually actuate as an alternative in the surveillance requirements. Additionally, the frequency of TS SRs 3.4.3.2 and 3.5.1.11 is being revised to be in accordance with the lST Program.
As a second option to performing a manual actuation of each SRV once per cycle, TVA is proposing, in accordance with the COR, to credit overlapping code and SR (testing) to ensure the capability of the SRV to open. Manual actuation testing will, however, be retained as an alternative in the TS SRs. The proposed revision to the SRs provides an alternative to the current requirement to demonstrate the capability of the relief valves to open when manually actuated during plant startup. This alternative provides another option to satisfy the SRs allowing a determination to be made that the valve is capable of being opened. Crediting of other testing and verification of electrical and pneumatic connections is in accordance with the COR, paragraph l-3410(d). The combination of testing the SRV actuator and solenoid valves and verifications of the capability of the SRV to open provides a complete verification of the functional capability in accordance with the COR. This testing is described in more detail below.
x The simulated automatic actuation test specified in SR 3.5.1.10 and additional surveillances associated with TS 3.3.5.1, Emergency Core Cooling System (ECCS)
Instrumentation, demonstrate the ability of various logics and controls to actuate the SRVs up to the point of energizing the solenoids. These tests are currently performed once per operating cycle and are unaffected by this proposed change.
x A solenoid valve (SOV) functional test will be performed in-situ for each SRV solenoid valve once per operating cycle. In the SOV functional test, a test rig with a pressure gauge will be connected downstream of the SOV pneumatic manifold in place of the SRV actuator. Each SOV will be energized, and pneumatic pressure at the downstream connection will be recorded and compared with pneumatic header pressure.
1 The BFN Units 1, 2, and 3 fifth ten-year IST interval is scheduled to begin on August 31, 2023. TVA has submitted a request to use a later code edition of the ASME OM Code (2020) for the fifth ten-year IST interval (ML22346A189), which will not impact the proposed changes in this license amendment request.
Enclosure CNL-23-018 E5 of 10 x
An SRV pilot air actuator functional test will be performed at an offsite test facility as part of certification testing for each SRV pilot assembly. The current practice of replacing all 13 SRV pilot assemblies each operating cycle will be maintained.
x SRV pilot setpoint testing is performed using steam at the offsite test facility as part of certification testing for each SRV pilot assembly, at intervals determined in accordance with the IST Program. This test is the existing test required by TS SR 3.4.3.1. In addition to demonstrating that the SRV pilot stage will actuate on high steam pressure in the safety mode, this test overlaps with the pilot assembly actuator functional test to demonstrate that the pilot stage will actuate in the relief mode.
x SRV main stage certification testing will be performed using steam at the offsite test facility at intervals determined in accordance with the IST Program. Main stage certification testing demonstrates that the main stage will open and port steam when actuated by the installed pilot stage.
x Receipt inspection is performed in accordance with the requirements of the TVA quality assurance program. The storage requirements in effect at BFN demonstrates that the valves are protected from exposure to the environment, airborne contamination, acceleration forces, and physical damage. Prior to installation, the valves are inspected for foreign material.
The COR allows a series of overlapping tests to individually test SRV components. Previous editions of the OM Code required that SRVs with auxiliary actuating devices that have been maintained or refurbished in place, removed for maintenance and testing, or both, and reinstalled be remotely actuated at reduced or normal system pressure to verify open and close capability of the valve before resumption of electric power generation. Paragraph I-3410(d) was revised in the 2004 edition of the ASME OM Code and no longer requires that SRVs be opened and closed at reduced or normal system pressure following maintenance. Rather, Paragraph I-3410(d) requires that each SRV that has been removed for maintenance or testing and reinstalled shall have the electrical and pneumatic connections verified either through mechanical/electrical inspection or test. The COR does not require that an SRV be tested as a unit. For example, the auxiliary actuating device can be tested independently of the main disk assembly. The test methods described above fully meet the requirements of the COR for pressure relief valves. Main stage certification testing demonstrates that the main stage will open and port steam when actuated by the installed pilot stage.
The proposed change revises the frequency for performing SR 3.4.3.2 and SR 3.5.1.11 to be in accordance with the IST Program. Specifying the required frequency through the IST Program is not a new technique and occurs throughout the TS (for example, SR 3.4.3.1 for verifying the safety function lift setpoints of the SRVs). Performing testing in accordance with the IST Program retains appropriate control over the testing methodology and specified frequency, because performance is required and is governed by 10 CFR 50.55a. Also, future OM Code changes could then be adopted without requiring a corresponding TS change, allowing NRC endorsed code changes to be more rapidly put in place. Additionally, this will allow crediting IST Program tests performed at frequencies other than the current 24 months as established by the SFCP.
Based on the above technical evaluation, TVA concludes that testing in accordance with the proposed changes to SR 3.4.3.2 and SR 3.5.1.11 will continue to demonstrate proper SRV operation without the need for in-situ testing with reactor steam and that this testing fully meets the requirements of the COR for safety and relief valves which the NRC staff has previously found to be acceptable.
Enclosure CNL-23-018 E6 of 10
4.0 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements Regulation 10 CFR 50.36, Technical specifications, provides the requirements for the content required in the TS. As stated in 10 CFR 50.36, the TSs include, among other things, Limiting Conditions for Operation (LCO) and SRs to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.
The proposed change revises BFN SR 3.4.3.2 and SR 3.5.1.11 by supplementing the current requirement to verify the SRVs open when manually actuated with an alternate requirement that verifies the SRVs are capable of being opened and by establishing the frequency of this testing to be in accordance with the IST Program. The proposed SR testing will continue to demonstrate proper SRV operation without the need for in-situ testing with reactor steam and that this testing fully meets the requirements of the COR for safety and relief valves.
Regulation 10 CFR 50.55a(f) requires that ASME Class 1, 2, and 3 components meet the requirements of the ASME OM Code, except where alternatives have been authorized pursuant to paragraphs (a)(3)(i) and (a)(3)(ii) of 10 CFR 50.55a. Currently, BFN Units 1, 2, and 3 SR 3.4.3.2 and SR 3.5.1.11 require that each main steam SRV must be manually actuated during a startup once reactor steam pressure and flow are adequate to perform the test.
Supplementing the manual actuation SRV test method in SR 3.4.3.2 and SR 3.5.1.11 with an alternative requirement to verify that the valves are capable of being opened as determined through a series of overlapping tests will continue to demonstrate proper SRV operation without the need for in-situ testing with reactor steam and that this testing fully meets the requirements of the COR for pressure relief valves.
The BFN units were designed and constructed based on the proposed General Design Criteria (GDC) published by the Atomic Energy Commission (AEC) in the Federal Register (32 FR 10213) on July 11, 1967 (hereafter called draft GDC). The AEC published the final rule that added Appendix A to 10 CFR Part 50, GDC for Nuclear Power Plants, in the Federal Register (36 FR 3255) on February 20, 1971 (hereafter called final GDC). As discussed in Appendix A of the Final Safety Analysis Report (FSAR), the licensee has made changes to the facility over the life of the plant that may have invoked the final GDC. The extent to which the final GDC have been invoked can be found in specific sections of the FSAR and in other design and licensing basis documentation. GDC 15, 35, and 37 are applicable to this LAR review.
Criterion 15Reactor coolant system design. The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.
Enclosure CNL-23-018 E7 of 10 Criterion 35Emergency core cooling. A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.
Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.
Criterion 37Testing of emergency core cooling system. The emergency core cooling system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.
TVA has evaluated the proposed changes against the applicable regulatory requirements and acceptance criteria. The analysis concludes that the proposed TS changes will continue to assure that the design and licensing requirements associated with the SRVs and their associated functions (i.e., safety, remote actuation, and ADS) are met. Based on this, there is reasonable assurance that the health and safety of the public, following approval of this TS change, are unaffected.
4.2 Precedent The proposed change is similar to NRC approved license amendments in References 2, 3, and 4. Prior approved amendments replaced manual actuation with verification of the capability to open. This proposed change provides both methods as acceptable means of verifying SRV functionality.
4.3 No Significant Hazards Consideration Determination Tennessee Valley Authority (TVA) proposes to revise the Browns Ferry Nuclear Plant (BFN)
Units 1, 2, and 3 Technical Specification (TS) Surveillance Requirement (SR) 3.4.3.2 and SR 3.5.1.11 by replacing the current requirement to verify the safety relief valves (SRVs) open when manually actuated with an alternate requirement that verifies the SRVs are capable of being opened. The proposed change also revises the frequency for these SRs to be in accordance with the Inservice Testing (IST) Program.
TVA has concluded that these changes do not involve a significant hazards consideration.
TVAs conclusion is based on its evaluation in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.91(a)(1) of the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below.
Enclosure CNL-23-018 E8 of 10
- 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed change does not involve a physical alteration of the plant (that is, no new or different type of equipment will be installed). The proposed change revises SR 3.4.3.2 and SR 3.5.1.11 by supplementing the current requirement to verify the SRVs open when manually actuated with an alternate requirement that verifies the SRVs are capable of being opened through a series of overlapping tests and requires the testing to be completed on a frequency in accordance with the IST Program. The proposed SR testing will continue to demonstrate proper SRV operation without the need for in-situ testing with reactor steam. This testing fully meets the requirements of the American Society of Mechanical Engineers (ASME)
Code for Operation and Maintenance of Nuclear Power Plants (OM Code) for safety and relief valves. Performing testing in accordance with the IST Program retains appropriate legal control over the testing methodology and specified frequency, since performance is required and governed by a code adopted into the regulation (i.e., 10 CFR 50.55a). Therefore, the proposed change does not adversely affect the ability of structures, systems, and components to perform their intended safety function to mitigate the consequences of event. Further, the proposed change does not increase the types and the amounts of radioactive effluent that may be released, nor significantly increase individual or cumulative occupation/public radiation exposures.
Therefore, the proposed changes do not involve a significant increase in the probability of consequences of an accident previously identified.
- 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No The proposed change revises SR 3.4.3.2 and SR 3.5.1.11 by supplementing the current requirement to verify the SRVs open when manually actuated with an alternate requirement that verifies the SRVs are capable of being opened through a series of overlapping tests and updating the frequency to be in accordance with the IST Program. It does not require any modification to the plant, and it does not alter the design configuration, or method of operation of plant equipment beyond its normal functional capabilities. The proposed change will not introduce failure modes that could result in a new accident, and the change does not alter assumptions made in the safety analysis.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
Enclosure CNL-23-018 E9 of 10
- 3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No The proposed change revises SR 3.4.3.2 and SR 3.5.1.11 by supplementing the current requirement to verify the SRVs open when manually actuated with an alternate requirement that verifies the SRVs are capable of being opened through a series of overlapping tests and updating the frequency to be in accordance with the IST Program. The proposed SR testing will continue to demonstrate proper SRV operation without the need for in-situ testing with reactor steam. It does not alter or exceed a design basis or safety limit. There is no change being made to safety analysis assumptions or the safety limits that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by the proposed change and the applicable requirements of 10 CFR 50.36(c)(3) will continue to be met.
Therefore, based on the above discussion, these proposed changes do not involve a reduction in the margin of safety.
Based on the above, it is concluded that the proposed amendment does not involve a 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.
4.4 Conclusions 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.
5.0 ENVIRONMENTAL CONSIDERATION
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 significant increase in the amounts of any radioactive 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-23-018 E10 of 10
6.0 REFERENCES
- 1. BWROG letter to NRC, BWR Owners Group Evaluations of NUREG-0737 Requirements II.K.3.16 and II.K.3.18, dated March 31, 1981 (8104200300)
- 2. NRC letter to Duke Energy Progress, LLC, Brunswick Steam Electric Plant, Units 1 and 2 - Issuance of Amendment Nos. 297 and 325 to Modify Technical Specification Surveillance Requirements 3.4.3.2 and 3.5.1.11 Regarding Safety Relief Valves (EPID L-2019-LLA-0043), dated January 8, 2020 (ML19316B057)
- 3. NRC letter to Detroit Edison Company, Fermi 2 - Issuance of Amendment to Modify Technical Specification Surveillance Requirements for Safety Relief Valves (TAC No. ME7829), dated December 21, 2012 (ML12321A234)
- 4. NRC letter to Northern States Power Company - Minnesota (NSPM), Monticello Nuclear Generating Plant - Issuance of Amendment Re: Testing of Main Steam Safety/Relief Valves (TAC No. ME7920), dated July 27, 2012 (ML12185A216)
Enclosure CNL-23-018 Proposed Technical Specification Changes (Markups) for BFN Units 1, 2, and 3 (6 pages)
S/RVs 3.4.3 BFN-UNIT 1 3.4-8 Amendment No. 234, 263, 269, 301, 315 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 Surveillance Frequency Control Program INSERVICE TESTING PROGRAM
ECCS - Operating 3.5.1 BFN-UNIT 1 3.5-7 Amendment No. 234, 254, 263, 315 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.1.10
NOTE-------------------------
Valve actuation may be excluded.
Verify the ADS actuates on an actual or simulated automatic initiation signal.
In accordance with the Surveillance Frequency Control Program SR 3.5.1.11
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 ADS valve is capable of being opened in accordance with the INSERVICE TESTING PROGRAM.
OR Verify each ADS valve opens when manually actuated.
In accordance with the Surveillance Frequency Control Program INSERVICE TESTING PROGRAM SR 3.5.1.12 (Deleted).
S/RVs 3.4.3 BFN-UNIT 2 3.4-8 Amendment No. 255, 325, 338 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 Surveillance Frequency Control Program INSERVICE TESTING PROGRAM
ECCS - Operating 3.5.1 BFN-UNIT 2 3.5-7 Amendment No. 255, 309, 338 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.1.10
NOTE-------------------------
Valve actuation may be excluded.
Verify the ADS actuates on an actual or simulated automatic initiation signal.
In accordance with the Surveillance Frequency Control Program SR 3.5.1.11
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 ADS valve is capable of being opened in accordance with the INSERVICE TESTING PROGRAM.
OR Verify each ADS valve opens when manually actuated.
In accordance with the Surveillance Frequency Control Program INSERVICE TESTING PROGRAM SR 3.5.1.12 (Deleted)
S/RVs 3.4.3 BFN-UNIT 3 3.4-8 Amendment No. 215, 285, 298 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 Surveillance Frequency Control Program INSERVICE TESTING PROGRAM
ECCS - Operating 3.5.1 BFN-UNIT 3 3.5-7 Amendment No. 215, 268, 298 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.1.10
NOTE-------------------------
Valve actuation may be excluded.
Verify the ADS actuates on an actual or simulated automatic initiation signal.
In accordance with the Surveillance Frequency Control Program SR 3.5.1.11
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 ADS valve is capable of being opened in accordance with the INSERVICE TESTING PROGRAM.
OR Verify each ADS valve opens when manually actuated.
In accordance with the Surveillance Frequency Control Program INSERVICE TESTING PROGRAM SR 3.5.1.12 (Deleted) 3
Enclosure CNL-23-018 Proposed Technical Specification Changes (Final Typed) for BFN Units 1, 2, and 3 (6 pages)
S/RVs 3.4.3 BFN-UNIT 1 3.4-8 Amendment No. 234, 263, 269, 301, 315, ___
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
,16(59,&(7(67,1*352*5$0.
OR Verify each required S/RV opens when manually actuated.
In accordance with the INSERVICE TESTING PROGRAM
ECCS - Operating 3.5.1 BFN-UNIT 1 3.5-7 Amendment No. 234, 254, 263, 315, ___
SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.1.10
NOTE-------------------------
Valve actuation may be excluded.
Verify the ADS actuates on an actual or simulated automatic initiation signal.
In accordance with the Surveillance Frequency Control Program SR 3.5.1.11 In accordance with the INSERVICE TESTING PROGRAM
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 ADS valve is capable of being opened in accordance with the I16(59,&(
7(67,1*352*5$0.
OR Verify each ADS valve is capable of being opened.
SR 3.5.1.12 (Deleted).
S/RVs 3.4.3 BFN-UNIT 2 3.4-8 Amendment No. 255, 325, 338, ___
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%.
SR 3.4.3.2 In accordance with the INSERVICE TESTING PROGRAM
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
,16(59,&(7(67,1*352*5$0.
OR Verify each required S/RV is capable of being opened.
ECCS - Operating 3.5.1 BFN-UNIT 2 3.5-7 Amendment No. 255, 309, 338, ___
SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.1.10
NOTE-------------------------
Valve actuation may be excluded.
Verify the ADS actuates on an actual or simulated automatic initiation signal.
In accordance with the Surveillance Frequency Control Program SR 3.5.1.11 In accordance with the INSERVICE TESTING PROGRAM
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 ADS valve is capable of being opened in accordance with the,16(59,&(
7(67,1*352*5$0.
OR Verify each ADS valve is capable of being opened.
SR 3.5.1.12 (Deleted)
S/RVs 3.4.3 BFN-UNIT 3 3.4-8 Amendment No. 215, 285, 298, ___
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%.
SR 3.4.3.2 In accordance with the INSERVICE TESTING PROGRAM
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
,16(59,&(7(67,1*352*5$0.
OR Verify each required S/RV is capable of being opened.
ECCS - Operating 3.5.1 BFN-UNIT 3 3.5-7 Amendment No. 215, 268, 298, ___
SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.1.10
NOTE-------------------------
Valve actuation may be excluded.
Verify the ADS actuates on an actual or simulated automatic initiation signal.
In accordance with the Surveillance Frequency Control Program SR 3.5.1.11 In accordance with the INSERVICE TESTING PROGRAM
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 ADS valve is capable of being opened in accordance with the,16(59,&(
7(67,1*352*5$0.
OR Verify each ADS valve is capable of being opened.
SR 3.5.1.12 (Deleted) 3
Enclosure CNL-23-018 Proposed Technical Specification Bases Changes (Markups) for BFN Unit 1 (For Information Only) (7 pages)
S/RVs B 3.4.3 (continued)
BFN-UNIT 1 B 3.4-21 Revision 0 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:
x Each S/RV main stage opens and passes steam when the associated pilot stage actuates; x Each S/RV pilot stage actuates to open the associated main stage when the pneumatic actuator is pressurized; x Each S/RV solenoid valves ports pneumatic pressure to the associated S/RV actuator when energized; and x 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.
S/RVs B 3.4.3 (continued)
BFN-UNIT 1 B 3.4-22 Revision 0 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 3.4-23 Revision 0, 43, 81, 123 March 22, 2020 BASES SURVEILLANCE SR 3.4.3.2 (continued)
REQUIREMENTS (continued)
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program 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.
ECCS - Operating B 3.5.1 (continued)
BFN-UNIT 1 B 3.5-20 Revision 0 BASES SURVEILLANCE SR 3.5.1.11 REQUIREMENTS (continued)
This Surveillance verifies that each ADS valve 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. 16). Proper ADS valve 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:
x Each ADS valve main stage opens and passes steam when the associated pilot stage actuates; x Each ADS valve pilot stage actuates to open the associated main stage when the pneumatic actuator is pressurized; x Each ADS valve solenoid valves ports pneumatic pressure to the associated S/RV actuator when energized; and x Each ADS valve 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.
ECCS - Operating B 3.5.1 (continued)
BFN-UNIT 1 B 3.5-21 Revision 0 BASES SURVEILLANCE SR 3.5.1.11 (continued)
REQUIREMENTS (continued)
Method 2 A manual actuation of each ADS valve is performed to verify that the valve and solenoid are functioning properly and that no blockage exists in the S/RV discharge lines. This is demonstrated by the response of the turbine control or bypass valve or by a change in the measured 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 ADS valves divert steam flow upon opening. Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this SR. Adequate pressure at which this SR 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. Reactor startup is allowed prior to performing this SR because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME 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 and provides adequate time to complete the Surveillance. SR 3.5.1.10 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.
ECCS - Operating B 3.5.1 (continued)
BFN-UNIT 1 B 3.5-22 Revision 0, 33, 43, 123 March 22, 2020 BASES SURVEILLANCE SR 3.5.1.11 (continued)
REQUIREMENTS (continued)
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program 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.
RPV Water Inventory Control B 3.5.2 BFN-UNIT 1 B 3.5-23 Revision 0, 125 January 26, 2022 BASES (continued)
REFERENCES
- 1. FSAR, Section 6.4.3.
- 2. FSAR, Section 6.4.4.
- 3. FSAR, Section 6.4.1.
- 4. FSAR, Section 6.4.2.
- 5. FSAR, Section 14.6.3.
- 6. FSAR, Section 14.6.5.
- 8. FSAR, Section 6.5.3.
- 9. 10 CFR 50.46.
- 11. Memorandum from R. L. Baer (NRC) to V. Stello, Jr. (NRC),
"Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.
- 12. 10 CFR 50, Appendix A, GDC 34, GDC 35, GDC 36, and GDC 37.
- 13. ANP-3377P Revision 3, Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU), August 2015.
- 14. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 15. GE-NE-B13-01755-2, "Relaxation of ECCS Parameters, Revision 2, December 1996.