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| | number = ML17047A610 | | | number = ML17047A610 |
| | issue date = 03/02/2017 | | | issue date = 03/02/2017 |
| | title = Beaver Valley Power Station, Unit No. 1 - Relief from the Requirements of the American Society of Mechanical Engineers Code (CAC No. MF8531) | | | title = Relief from the Requirements of the American Society of Mechanical Engineers Code |
| | author name = Koenick S S | | | author name = Koenick S |
| | author affiliation = NRC/NRR/DORL/LPLI | | | author affiliation = NRC/NRR/DORL/LPLI |
| | addressee name = Richey M L | | | addressee name = Richey M |
| | addressee affiliation = FirstEnergy Nuclear Operating Co | | | addressee affiliation = FirstEnergy Nuclear Operating Co |
| | docket = 05000334 | | | docket = 05000334 |
| | license number = DPR-066 | | | license number = DPR-066 |
| | contact person = Taylor L A, NRR/DORL/LPL1, 415-7128 | | | contact person = Taylor L, NRR/DORL/LPL1, 415-7128 |
| | case reference number = CAC MF8531 | | | case reference number = CAC MF8531 |
| | document type = Code Relief or Alternative, Letter, Safety Evaluation | | | document type = Code Relief or Alternative, Letter, Safety Evaluation |
| | page count = 8 | | | page count = 8 |
| | project = CAC:MF8531 | | | project = CAC:MF8531 |
| | | stage = Approval |
| }} | | }} |
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| =Text= | | =Text= |
| {{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Marty L. Richey, Site Vice President FirstEnergy Nuclear Operating Company Beaver Valley Power Station Mail Stop A-BV-SEB1 P.O. Box 4, Route 168 Shippingport, PA 15077 March 2, 2017 SUBJECT: BEAVER VALLEY POWER STATION, UNIT NO. 1 -RELIEF FROM THE REQUIREMENTS OF THE ASME CODE (CAC NO. MF8531) Dear Mr. Richey: By letter dated October 24, 2016, as supplemented by letter dated January 13, 2017 (Agencywide Documents Access and Management System Accession Nos. ML 16298A289 and ML 17013A483, respectively), FirstEnergy Nuclear Operating Company (FENOC or the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to certain American Society of Mechanical Engineers (ASME) Code for Operating and Maintenance of Nuclear Power Plants (OM Code) requirements at Beaver Valley Power Station, Unit No. 1. Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee requested to use the proposed alternative on the basis that complying with the specified requirement would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. The proposed alternative is to delay a relief valve test to the maintenance and refueling outage of the spring of 2018. The NRC staff has reviewed the subject request and determined, as set forth in the enclosed safety evaluation, that FENOC has adequately addressed all of the regulatory requirements set for in 10 CFR 50.55a and that the proposed alternative provides reasonable assurance that the affected component is operationally ready. Accordingly, the NRC staff concludes that FENOC has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2) and is in compliance with ASME OM Code requirements. Therefore, the NRC staff authorizes the proposed alternative request to test valve RV-1 RH-721 during the next refueling outage, which is currently scheduled for the spring of 2018. All other ASME OM Code requirements for which relief was not specifically requested and approved in this relief request remain applicable. | | {{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 2, 2017 Mr. Marty L. Richey, Site Vice President FirstEnergy Nuclear Operating Company Beaver Valley Power Station Mail Stop A-BV-SEB1 P.O. Box 4, Route 168 Shippingport, PA 15077 |
| M. Richey -2 -If you have any questions, please contact the Project Manager, Taylor Lamb, at 301-415-7128 or Taylor.Lamb@nrc.gov. Docket No. 50-334 Enclosure: Safety Evaluation cc w/encl: Distribution via Listserv Sincerely, Stephen S. Koenick, Acting Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation REQUEST FOR ALTERNATIVE VRR6 REGARDING RESIDUAL HEAT REMOVAL SYSTEM VALVE TESTING FIRSTENERGY NUCLEAR OPERATING COMPANY BEAVER VALLEY POWER STATION. UNIT NO. 1 DOCKET NO. 50-334 1.0 INTRODUCTION By letter dated October 24, 2016, as supplemented by letter dated January 13, 2017 (Agencywide Documents Access and Management System Accession Nos. ML 16298A289 and ML 17013A483, respectively), FirstEnergy Nuclear Operating Company (FE NOC or the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC or the Commission) for the use of alternatives to certain American Society of Mechanical Engineers (ASME) Code for Operating and Maintenance of Nuclear Power Plants (OM Code) requirements at Beaver Valley Power Station, Unit No. 1 (BVPS-1). Specifically, pursuant to Title 1 O of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee requested to use the proposed alternative on the basis that complying with the specified requirement would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. The proposed alternative is to delay a relief valve test to the maintenance and refueling outage of the spring of 2018. 2.0 REGULATORY EVALUATION The regulations in 10 CFR 50.55a(f), "lnservice testing requirements," require, in part, that inservice testing (IST) of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda, except where alternatives have been authorized pursuant to 1 O CFR 50.55a(z)(1) or 1 O CFR 50.55a(z)(2). In proposing alternatives, a licensee must demonstrate that the proposed alternatives provide an acceptable level of quality and safety (10 CFR 50.55a(z)(1)), or compliance would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety ( 10 CFR 50.55a(z)(2) ). Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request, and the Commission to authorize, the alternative requested by the licensee. Enclosure | | |
| -2 -3.0 TECHNICAL EVALUATION 3.1 Licensee Relief Request No. VRR6 Alternative testing is requested for the BVPS-1 residual heat removal (RHR) pump relief valve RV-1 RH-721 (Class 2, Category C). Applicable ASME OM Code requirements, as stated by the licensee, are: Mandatory Appendix I, 1-1350, "Test Frequency, Class 2 and 3 Pressure Relief Valves," Paragraph (c), "Requirements for Testing Additional Valves," states in part: Additional valves shall be tested in accordance with the following requirements: 1. For each valve tested for which as-found set-pressure (first test actuation) exceeds the greater of either the +/- tolerance limit of the Owner established set-pressure acceptance criteria of 1-131 O(e) or +/-3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group. The licensee states: Reason for Request Charging system letdown relief valve RV-1 CH-203 has a cold differential test pressure (CDTP) of 606 pounds per square inch gauge (psig) with an upper acceptable limit of plus 3 percent or 624 psig. The CDTP is a temperature compensated test pressure used to account for the difference between the ambient temperature at the test stand and the higher temperature at the tested valve's installed location in the plant. During the 24th Beaver Valley Power Station, Unit No. 1 (BVPS-1 ), maintenance and refueling outage (1 R24) when RV-1 CH-203 was tested on the test stand at ambient conditions, the relief valve lifted at 631.7 psig, which was 4.25 percent above its CDTP. This exceeded the plus 3 percent limit specified in Paragraph l-1350(c)(1) requiring the only other valve in the sample group, RV-1 RH-721, to be tested. However, because the ASME OM Code, Appendix I, Paragraph l-1350(c)(1) allows for the owner to establish set-pressure acceptance criteria (a plus or minus tolerance limit), a limit of plus 5 percent was calculated after the valve exceeded the plus 3 percent limit specified in the ASME OM Code. Relief valve RV-1 CH-203 test results were within the plus 5 percent acceptance criteria. Therefore, valve RV-1 RH-721 was not tested. When it was determined that FENOC was not permitted to provide an owner specified limit after the valve was tested, the plant lineup was being established for plant startup with the reactor core fully loaded with fuel. Testing of RV-1 RH-721 cannot be performed in-place. The relief valve must be removed from the system and tested on a test stand. In order to remove and test RV-1 RH-721, the RHR system would have to be shut down and the entire
| | ==SUBJECT:== |
| -3 -system drained. This is normally performed when the RHR system is not required to be in operation, which is when fuel is removed from the reactor core. Based on the plant lineup and the need to remove the valve from the system to perform the test, the sample size could not be expanded as required and is not in compliance with the ASME OM Code. Therefore, a delay is proposed to test RV-1 RH-721 during the 25th BVPS-1 maintenance and refueling outage (1 R25) when fuel is removed from the core (1 R25 is scheduled for the spring of 2018). To unload the reactor core in order to test RV-1 RH-721 provides a hardship without a compensating increase in quality and safety. Proposed Alternative and Basis for Use The proposed alternative is to delay testing of relief valve RV-1 RH-721 until 1 R25. At that time, the RHR system can be drained, and RV-1 RH-721 can be tested. Valve RV-1 RH-721 is considered operationally ready until tested during 1 R25 for the following reasons: 1) A work order history review shows that two different serial-numbered (S/N) valves have been swapped in and out of the installed location for RV-1-RH-721 as follows: 1 R 11 (April 1996) S/N 53237M 1 installed (set at 610 psig) 1R13 (March 2000) S/N 53237M1 removed (as-found set at 616 psig), and S/N N69973-01-0001 installed (set at 608 psig) 1R15 (March 2003) S/N N69973-01-0001 removed (as-found set at 51 O psig due to leakage), and S/N 53237M1 installed (set at 610 psig) 1R17 (January 2006) S/N 53237M1 removed (as-found set at 603 psig), and SIN N69973-01-0001 installed (set at 610 psig) 1 R20 (October 2010) S/N N69973-01-0001 removed (as-found set at 611.2 psig), and S/N 53237M1 installed (set at 606 psig) The setpoint testing listed above was performed on a test stand at ambient conditions. The preceding data shows that valve S/N N69973-01-0001 has drifted high approximately 2 psig over a 3 to 4 year period while valve S/N 53237M1 has drifted high approximately 6 psig over a 4 year period. If the approximate 6 psig
| | BEAVER VALLEY POWER STATION, UNIT NO. 1 -RELIEF FROM THE REQUIREMENTS OF THE ASME CODE (CAC NO. MF8531) |
| -4 -high drift for valve S/N 53237M1 is extrapolated over a 7.5 year period from 1R20 to 1R25, RV-1RH-721 would be expected to lift no higherthan 618 psig. Therefore, based on past performance, there is reasonable assurance that RV-1 RH-721 would not lift greater than plus 3 percent above its set-pressure when tested during the next maintenance and refueling outage. 2) Other than a three-year period from 1R13 to 1R15 when the valve was found to lift low due to leakage, RV-1RH-721 has shown a history of consistent performance. Therefore, there is reasonable assurance that valve RV-1 RH-721 will continue to be operationally ready until the next scheduled test during 1 R25. Further, the test interval from 1 R20 to 1 R25 (7.5 years) is conservatively shorter than the maximum 10-year test interval requirement of ASME OM Code, Mandatory Appendix I, Paragraph l-1350(a). In addition, during operation at power, a surveillance verification log monitors annunciator window A 1-125, "Residual Heat Removal Pump Discharge Pressure HIGH." This. annunciator alarms if pressure reaches 550 psig and provides corrective actions to take in accordance with an alarm response procedure. The corrective actions would relieve pressure before it reaches the RV-1 RH-721 set-pressure of 600 psig. The proposed alternative to ASME OM Code, Mandatory Appendix I, paragraph l-1350(c) would delay the test of relief valve RV-1 RH-721 until 1 R25. Therefore, there is reasonable assurance that relief valve RV-1 RH-721 will continue to be operationally ready until 1 R25. The proposed alternative is requested for use during the remainder of the fourth 10-year IST interval. 3.2 NRC Staff Evaluation ASME OM Code, 2001 Edition through 2003 Addenda, Section Mandatory Appendix I, "lnservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants," paragraph l-1350(a), 10-year test Interval states that: Class 2 and 3 pressure relief valves, with the exception of PWR [pressurized-water reactor] main steam safety valves, shall be tested every 10 years, starting with initial electric power generation. No maximum limit is specified for the number of valves to be tested during any single plant operating cycle; however, a minimum of 20% of the valves from each valve group shall be tested within any 48-month interval. This 20% shall consist of valves that have not been tested during the current 10-year test interval, if they exist. Requirements for testing additional valves in a group is detailed in paragraph l-1350(c)(1), which states that, "For each valve tested for which the as-found set-pressure (first test actuation) exceeds the greater of either the+/- tolerance limit of the Owner established set-pressure acceptance criteria of l-1310(e) or +/-3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group."
| | |
| -5 -As noted in the alternative request, relief valve RV-1 CH-203 was recently tested pursuant to the required ASME OM Code test interval and failed to meet the standard ASME OM Code tolerance acceptance value of +/-3 percent of valve nameplate set-pressure. The standard ASME OM Code tolerance acceptance value was used because the licensee did not have an owner-established set-pressure acceptance criteria specified. The valve failure now required additional valves of the valve group to be tested. A valve group is defined in ASME OM Code, 2001 Edition through 2003 Addenda, as "Valves of the same manufacturer, type, system application, and service media." Valve RV-1 CH-203 has only one other valve associated with it in its group, which is RV-1 RH-721. Relief valve RV-1 RH-721 protects the RHR system. The valve cannot be tested in situ and must be removed from the system. To remove the valve from the system will require draining the RHR system, which is normally performed when the fuel is removed from the reactor core. The next scheduled fuel removal is the spring of 2018. To perform the required ASME OM Code test now would represent a hardship, without a compensating increase in quality and safety. The licensee proposes to delay testing RV-1 RH-721 until the next refuel outage when the RHR system can be removed from service after the reactor core is off-loaded. The NRC staff reviewed the performance history of the setpoint testing. Considering the sample, in which there was no instrument drift high above 1 percent, the NRC staff has reasonable assurance that the valve would not lift greater than +3 percent. Furthermore, the NRC staff credits the fact that at power, a surveillance verification log monitors annunciator window A 1-125 and that the annunciator alarms if pressure reaches 550 psig. The resulting corrective actions would relieve pressure before they reach the RV-1 RH-721 set-pressure of 600 psig. Therefore, the NRC staff concludes that the proposed alternative is acceptable. 4.0 CONCLUSION As set forth above, the NRC staff determined that the proposed alternative provides reasonable assurance that the affected component is operationally ready. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(z)(2). All other ASME OM Code requirements for which relief was not specifically requested and approved in the subject requests for relief remain applicable. Therefore, the NRC staff authorizes the BVPS-1 proposed alternative request to test valve RV-1 RH-721 during the next refueling outage, which is currently scheduled for the spring of 2018. Principal Contributor: M. Farnan Date: March 2, 2017 M. Richey -3 -SUBJECT: BEAVER VALLEY POWER STATION, UNIT NO. 1 -RELIEF FROM THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS CODE (CAC NO. MF8531) DATED MARCH 2, 2017 DISTRIBUTION: Public LPLI R/F RidsRgn1 MailCenter Resource RidsACRS_MailCTR Resource RidsNrrDorlLpl1 Resource RidsNrrPMBeaverValley Resource RidsNrrLALRonewicz Resource RidsNrrDeEpnb Resource MFarnan, NRR JBowen, OEDO ADAMS Accession No.: ML 17047A610 OFFICE DORL/LP LI/PM DORL/LPLl/LA NAME TLamb LRonewicz DATE 02/21/2017 02/21/2017 *by e-mail dated DE/EPNB/BC* DORL/LP LI/BC DAiiey SKoenick 02/13/2017 02/28/2017 OFFICIAL RECORD COPY DORL/LP LI/PM TLamb 03/02/2017
| | ==Dear Mr. Richey:== |
| }} | | |
| | By letter dated October 24, 2016, as supplemented by letter dated January 13, 2017 (Agencywide Documents Access and Management System Accession Nos. ML16298A289 and ML17013A483, respectively), FirstEnergy Nuclear Operating Company (FENOC or the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to certain American Society of Mechanical Engineers (ASME) Code for Operating and Maintenance of Nuclear Power Plants (OM Code) requirements at Beaver Valley Power Station, Unit No. 1. |
| | Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee requested to use the proposed alternative on the basis that complying with the specified requirement would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. The proposed alternative is to delay a relief valve test to the maintenance and refueling outage of the spring of 2018. |
| | The NRC staff has reviewed the subject request and determined, as set forth in the enclosed safety evaluation, that FENOC has adequately addressed all of the regulatory requirements set for in 10 CFR 50.55a and that the proposed alternative provides reasonable assurance that the affected component is operationally ready. Accordingly, the NRC staff concludes that FENOC has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2) and is in compliance with ASME OM Code requirements. Therefore, the NRC staff authorizes the proposed alternative request to test valve RV-1 RH-721 during the next refueling outage, which is currently scheduled for the spring of 2018. |
| | All other ASME OM Code requirements for which relief was not specifically requested and approved in this relief request remain applicable. |
| | |
| | M. Richey If you have any questions, please contact the Project Manager, Taylor Lamb, at 301-415-7128 or Taylor.Lamb@nrc.gov. |
| | Sincerely, |
| | ~j)(A Stephen S. Koenick, Acting Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-334 |
| | |
| | ==Enclosure:== |
| | |
| | Safety Evaluation cc w/encl: Distribution via Listserv |
| | |
| | REQUEST FOR ALTERNATIVE VRR6 REGARDING RESIDUAL HEAT REMOVAL SYSTEM VALVE TESTING FIRSTENERGY NUCLEAR OPERATING COMPANY BEAVER VALLEY POWER STATION. UNIT NO. 1 DOCKET NO. 50-334 |
| | |
| | ==1.0 INTRODUCTION== |
| | |
| | By letter dated October 24, 2016, as supplemented by letter dated January 13, 2017 (Agencywide Documents Access and Management System Accession Nos. ML16298A289 and ML17013A483, respectively), FirstEnergy Nuclear Operating Company (FE NOC or the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC or the Commission) for the use of alternatives to certain American Society of Mechanical Engineers (ASME) Code for Operating and Maintenance of Nuclear Power Plants (OM Code) requirements at Beaver Valley Power Station, Unit No. 1 (BVPS-1). |
| | Specifically, pursuant to Title 1O of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee requested to use the proposed alternative on the basis that complying with the specified requirement would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. The proposed alternative is to delay a relief valve test to the maintenance and refueling outage of the spring of 2018. |
| | |
| | ==2.0 REGULATORY EVALUATION== |
| | |
| | The regulations in 10 CFR 50.55a(f), "lnservice testing requirements," require, in part, that inservice testing (IST) of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda, except where alternatives have been authorized pursuant to 10 CFR 50.55a(z)(1) or 10 CFR 50.55a(z)(2). |
| | In proposing alternatives, a licensee must demonstrate that the proposed alternatives provide an acceptable level of quality and safety (10 CFR 50.55a(z)(1)), or compliance would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety |
| | ( 10 CFR 50.55a(z)(2) ). |
| | Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request, and the Commission to authorize, the alternative requested by the licensee. |
| | Enclosure |
| | |
| | ==3.0 TECHNICAL EVALUATION== |
| | |
| | 3.1 Licensee Relief Request No. VRR6 Alternative testing is requested for the BVPS-1 residual heat removal (RHR) pump relief valve RV-1 RH-721 (Class 2, Category C). |
| | Applicable ASME OM Code requirements, as stated by the licensee, are: |
| | Mandatory Appendix I, 1-1350, "Test Frequency, Class 2 and 3 Pressure Relief Valves," Paragraph (c), "Requirements for Testing Additional Valves," states in part: |
| | Additional valves shall be tested in accordance with the following requirements: |
| | : 1. For each valve tested for which as-found set-pressure (first test actuation) exceeds the greater of either the +/- tolerance limit of the Owner established set-pressure acceptance criteria of 1-131 O(e) or +/-3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group. |
| | The licensee states: |
| | |
| | ===Reason for Request=== |
| | Charging system letdown relief valve RV-1 CH-203 has a cold differential test pressure (CDTP) of 606 pounds per square inch gauge (psig) with an upper acceptable limit of plus 3 percent or 624 psig. The CDTP is a temperature compensated test pressure used to account for the difference between the ambient temperature at the test stand and the higher temperature at the tested valve's installed location in the plant. During the 24th Beaver Valley Power Station, Unit No. 1 (BVPS-1 ), maintenance and refueling outage (1 R24) when RV-1 CH-203 was tested on the test stand at ambient conditions, the relief valve lifted at 631.7 psig, which was 4.25 percent above its CDTP. This exceeded the plus 3 percent limit specified in Paragraph l-1350(c)(1) requiring the only other valve in the sample group, RV-1 RH-721, to be tested. However, because the ASME OM Code, Appendix I, Paragraph l-1350(c)(1) allows for the owner to establish set-pressure acceptance criteria (a plus or minus tolerance limit), a limit of plus 5 percent was calculated after the valve exceeded the plus 3 percent limit specified in the ASME OM Code. Relief valve RV-1 CH-203 test results were within the plus 5 percent acceptance criteria. Therefore, valve RV-1 RH-721 was not tested. When it was determined that FENOC was not permitted to provide an owner specified limit after the valve was tested, the plant lineup was being established for plant startup with the reactor core fully loaded with fuel. |
| | Testing of RV-1 RH-721 cannot be performed in-place. The relief valve must be removed from the system and tested on a test stand. In order to remove and test RV-1 RH-721, the RHR system would have to be shut down and the entire |
| | |
| | system drained. This is normally performed when the RHR system is not required to be in operation, which is when fuel is removed from the reactor core. |
| | Based on the plant lineup and the need to remove the valve from the system to perform the test, the sample size could not be expanded as required and is not in compliance with the ASME OM Code. Therefore, a delay is proposed to test RV-1 RH-721 during the 25th BVPS-1 maintenance and refueling outage (1 R25) when fuel is removed from the core (1 R25 is scheduled for the spring of 2018). |
| | To unload the reactor core in order to test RV-1 RH-721 provides a hardship without a compensating increase in quality and safety. |
| | Proposed Alternative and Basis for Use The proposed alternative is to delay testing of relief valve RV-1 RH-721 until 1R25. At that time, the RHR system can be drained, and RV-1 RH-721 can be tested. |
| | Valve RV-1 RH-721 is considered operationally ready until tested during 1R25 for the following reasons: |
| | : 1) A work order history review shows that two different serial-numbered (S/N) valves have been swapped in and out of the installed location for RV-1-RH-721 as follows: |
| | 1R 11 (April 1996) S/N 53237M 1 installed (set at 610 psig) 1R13 (March 2000) S/N 53237M1 removed (as-found set at 616 psig), |
| | and S/N N69973-01-0001 installed (set at 608 psig) 1R15 (March 2003) S/N N69973-01-0001 removed (as-found set at 51 O psig due to leakage), and S/N 53237M1 installed (set at 610 psig) 1R17 (January 2006) S/N 53237M1 removed (as-found set at 603 psig), |
| | and SIN N69973-01-0001 installed (set at 610 psig) 1R20 (October 2010) S/N N69973-01-0001 removed (as-found set at 611.2 psig), and S/N 53237M1 installed (set at 606 psig) |
| | The setpoint testing listed above was performed on a test stand at ambient conditions. |
| | The preceding data shows that valve S/N N69973-01-0001 has drifted high approximately 2 psig over a 3 to 4 year period while valve S/N 53237M1 has drifted high approximately 6 psig over a 4 year period. If the approximate 6 psig |
| | |
| | high drift for valve S/N 53237M1 is extrapolated over a 7.5 year period from 1R20 to 1R25, RV-1RH-721 would be expected to lift no higherthan 618 psig. |
| | Therefore, based on past performance, there is reasonable assurance that RV-1 RH-721 would not lift greater than plus 3 percent above its set-pressure when tested during the next maintenance and refueling outage. |
| | : 2) Other than a three-year period from 1R13 to 1R15 when the valve was found to lift low due to leakage, RV-1RH-721 has shown a history of consistent performance. Therefore, there is reasonable assurance that valve RV-1 RH-721 will continue to be operationally ready until the next scheduled test during 1R25. Further, the test interval from 1R20 to 1R25 (7.5 years) is conservatively shorter than the maximum 10-year test interval requirement of ASME OM Code, Mandatory Appendix I, Paragraph l-1350(a). |
| | In addition, during operation at power, a surveillance verification log monitors annunciator window A 1-125, "Residual Heat Removal Pump Discharge Pressure HIGH." This. annunciator alarms if pressure reaches 550 psig and provides corrective actions to take in accordance with an alarm response procedure. The corrective actions would relieve pressure before it reaches the RV-1 RH-721 set-pressure of 600 psig. |
| | The proposed alternative to ASME OM Code, Mandatory Appendix I, paragraph l-1350(c) would delay the test of relief valve RV-1 RH-721 until 1R25. |
| | Therefore, there is reasonable assurance that relief valve RV-1 RH-721 will continue to be operationally ready until 1R25. |
| | The proposed alternative is requested for use during the remainder of the fourth 10-year IST interval. |
| | 3.2 NRC Staff Evaluation ASME OM Code, 2001 Edition through 2003 Addenda, Section Mandatory Appendix I, "lnservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants," |
| | paragraph l-1350(a), 10-year test Interval states that: |
| | Class 2 and 3 pressure relief valves, with the exception of PWR |
| | [pressurized-water reactor] main steam safety valves, shall be tested every 10 years, starting with initial electric power generation. No maximum limit is specified for the number of valves to be tested during any single plant operating cycle; however, a minimum of 20% of the valves from each valve group shall be tested within any 48-month interval. This 20% shall consist of valves that have not been tested during the current 10-year test interval, if they exist. |
| | Requirements for testing additional valves in a group is detailed in paragraph l-1350(c)(1), |
| | which states that, "For each valve tested for which the as-found set-pressure (first test actuation) exceeds the greater of either the+/- tolerance limit of the Owner established set-pressure acceptance criteria of l-1310(e) or +/-3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group." |
| | |
| | As noted in the alternative request, relief valve RV-1 CH-203 was recently tested pursuant to the required ASME OM Code test interval and failed to meet the standard ASME OM Code tolerance acceptance value of +/-3 percent of valve nameplate set-pressure. The standard ASME OM Code tolerance acceptance value was used because the licensee did not have an owner-established set-pressure acceptance criteria specified. The valve failure now required additional valves of the valve group to be tested. A valve group is defined in ASME OM Code, 2001 Edition through 2003 Addenda, as "Valves of the same manufacturer, type, system application, and service media." Valve RV-1 CH-203 has only one other valve associated with it in its group, which is RV-1 RH-721. |
| | Relief valve RV-1 RH-721 protects the RHR system. The valve cannot be tested in situ and must be removed from the system. To remove the valve from the system will require draining the RHR system, which is normally performed when the fuel is removed from the reactor core. |
| | The next scheduled fuel removal is the spring of 2018. To perform the required ASME OM Code test now would represent a hardship, without a compensating increase in quality and safety. |
| | The licensee proposes to delay testing RV-1 RH-721 until the next refuel outage when the RHR system can be removed from service after the reactor core is off-loaded. The NRC staff reviewed the performance history of the setpoint testing. Considering the sample, in which there was no instrument drift high above 1 percent, the NRC staff has reasonable assurance that the valve would not lift greater than +3 percent. Furthermore, the NRC staff credits the fact that at power, a surveillance verification log monitors annunciator window A 1-125 and that the annunciator alarms if pressure reaches 550 psig. The resulting corrective actions would relieve pressure before they reach the RV-1 RH-721 set-pressure of 600 psig. Therefore, the NRC staff concludes that the proposed alternative is acceptable. |
| | |
| | ==4.0 CONCLUSION== |
| | |
| | As set forth above, the NRC staff determined that the proposed alternative provides reasonable assurance that the affected component is operationally ready. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(z)(2). |
| | All other ASME OM Code requirements for which relief was not specifically requested and approved in the subject requests for relief remain applicable. |
| | Therefore, the NRC staff authorizes the BVPS-1 proposed alternative request to test valve RV-1 RH-721 during the next refueling outage, which is currently scheduled for the spring of 2018. |
| | Principal Contributor: M. Farnan Date: March 2, 2017 |
| | |
| | ML17047A610 *by e-mail dated OFFICE DORL/LP LI/PM DORL/LPLl/LA DE/EPNB/BC* DORL/LP LI/BC DORL/LP LI/PM NAME TLamb LRonewicz DAiiey SKoenick TLamb DATE 02/21/2017 02/21/2017 02/13/2017 02/28/2017 03/02/2017}} |
Letter Sequence Approval |
|---|
|
|
MONTHYEARML16319A0572016-12-0101 December 2016
Acceptance of Requested Licensing Action Relief Request for Proposed Alternative to Perform As-Found Set-Pressure Test
Project stage: Acceptance Review
ML17047A6102017-03-0202 March 2017
Relief from the Requirements of the American Society of Mechanical Engineers Code
Project stage: Approval
ML17074A0192017-03-16016 March 2017
Acceptance of Requested Licensing Action Proposed Alternative to ASME Code Reactor Pressure Vessel Head Penetration Examination Frequency Requirements (Request 1-TYP-4-RV-05)
Project stage: Acceptance Review
2017-03-16
[Table View] |
|
|---|
Category:Code Relief or Alternative
MONTHYEARML24226A3652024-05-13013 May 2024
Approval of Request for Alternative from Certain Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code
ML23249A1842023-09-18018 September 2023
Authorization and Safety Evaluation for Alternative Request No. 2-TYP-4-RV-06
ML22082A2532022-03-29029 March 2022
Correction Letter - Issuance of Authorization of Proposed Alternative to Use ASME OM Code Case OMN-27
ML22012A2972022-01-21021 January 2022
Proposed Alternative to Use ASME OM Code Case OMN-27
L-20-256, Request to Use Provision in Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI2020-09-28028 September 2020
Request to Use Provision in Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI
ML20223A0142020-08-20020 August 2020
Issuance of Relief Request SSR-1, Revision 0, from the Requirements of the ASME Code (EPID L-2020-LLR-0050 (COVID-19))
ML20206K8652020-08-0707 August 2020
Relief Requests 2-TYP-4-RV-06 and 2-TYP-4-RV-07 from the Requirements of the ASME Code (EPID L-2020-LLR-0053 and EPID L-2020-LLR-0054 (COVID-19))
ML20145A0002020-06-0303 June 2020
Relief from the Requirements of the ASME Code for Requests VRR6 and VRR5 (EPID L-2020-LLR-0049 and EPID L-2020-LLR-0051 (COVID-19))
ML20100N3222020-04-0909 April 2020
Verbal Relief for Penetration Evaluation and Hot Leg Nozzles - Delivered 4/9/2020 at 10:00 Am
ML20099B2572020-04-0808 April 2020
Verbal Relief Unit 2 CIV
ML20098F3012020-04-0707 April 2020
Verbal Relief for Appendix I Safety Relief Valves
ML20095J2192020-04-0404 April 2020
Email Beaver Valley Power Station, Unit 2 - Verbal Relief for MOVs - Delivered 4/4/2020 at 4:00 Pm
ML20095J0992020-04-0404 April 2020
Email Beaver Valley Power Station, Unit 2 - Verbal Relief for Snubbers - Delivered 4/4/2020 at 4:00 P.M
L-20-117, 10 CFR 50.55a Request Number 2-TYP-4-RV-06, Hardship for Hot Leg Nozzle Inspections2020-04-0303 April 2020
10 CFR 50.55a Request Number 2-TYP-4-RV-06, Hardship for Hot Leg Nozzle Inspections
L-20-118, CFR 50.55a Request Number SRR-1, Revision 0, Snubber Testing2020-04-0303 April 2020
CFR 50.55a Request Number SRR-1, Revision 0, Snubber Testing
L-20-060, CFR 50.55a Request Number: VRR4, Revision 0, Containment Isolation Valve Test Frequency2020-04-0202 April 2020
CFR 50.55a Request Number: VRR4, Revision 0, Containment Isolation Valve Test Frequency
L-20-116, CFR 50.55a Request Number VRR6, Revision 0, Motor-Operated Valve Test Frequency2020-04-0101 April 2020
CFR 50.55a Request Number VRR6, Revision 0, Motor-Operated Valve Test Frequency
ML20079F8162020-03-26026 March 2020
Relief Requests 1-TYP-4-C2.21-1 and 1-TYP-4-RA-1 Regarding Weld Examination Coverage for the Fourth Inservice Inspection Interval
ML19275E2942019-10-16016 October 2019
Issuance of Relief Request Proposed Alternative to Reactor Vessel Nozzle Weld Examination Frequency Requirements in Lieu of Specific ASME Code Requirements
L-19-107, Impractical American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Examination Requirements2019-08-27027 August 2019
Impractical American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Examination Requirements
ML19051A1082019-02-20020 February 2019
Request Alternative Examination Frequency for Reactor Vessel Nozzle to Safe-End Welds (Request 2-TYP-4-RV-05, Revision 0)
ML18227A7332018-08-27027 August 2018
Request for Relief from the Requirements of the ASME Code
ML18004A1222018-01-22022 January 2018
FENOC-Beaver Valley, Davis-Besse, and Perry - Alternative for the Use of ASME Code Case N-513-4 (CAC Nos. MG0120, MG0121, MG0122, and MG0123; EPID L-2017-LLR-0088)
L-17-317, Request to Extend Certain Reactor Vessel Inspections from 10 to 20 Years (Request 1-TYP-4-BN-01)2017-11-15015 November 2017
Request to Extend Certain Reactor Vessel Inspections from 10 to 20 Years (Request 1-TYP-4-BN-01)
L-17-308, 10 CFR 50.55a Request for Alternate Reactor Vessel Nozzle Flaw Depth Sizing Criteria (Request 2-TYP-4-RVSE-2)2017-10-25025 October 2017
10 CFR 50.55a Request for Alternate Reactor Vessel Nozzle Flaw Depth Sizing Criteria (Request 2-TYP-4-RVSE-2)
ML17167A0672017-06-26026 June 2017
Requests for Alternatives and Requests for Relief Fourth 10-Year Inservice Test Program Interval (CAC Nos. MF8333-MF8356). Note: Correction Safety Evaluation See ML17255A526
ML17159A4422017-06-26026 June 2017
Requests for Alternatives and Requests for Relief Fifth 10-Year Inservice Testing Program Interval (CAC Nos. MF8332 Through MF8357). Note: Correction Safety Evaluation See ML17255A508
ML17047A6102017-03-0202 March 2017
Relief from the Requirements of the American Society of Mechanical Engineers Code
ML17041A1852017-03-0202 March 2017
Relief from the Requirements of the ASME Code
ML17048A0042017-03-0202 March 2017
Relief from the Requirements of the ASME Code
ML16328A1252017-01-23023 January 2017
Relief from the Requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) (TAC No. MF7780 - MF7783)
ML16190A1332016-12-27027 December 2016
Relief from the Requirements of the ASME Code
ML16319A0572016-12-0101 December 2016
Acceptance of Requested Licensing Action Relief Request for Proposed Alternative to Perform As-Found Set-Pressure Test
ML16257A6212016-11-21021 November 2016
Relief Request BV2-PZR-01, Regarding Alternative to Requirements for Components Connected to the Steam Side of the Pressurizer
ML16228A4082016-10-21021 October 2016
Correction to Relief Request 2-TYP-3-RV-04, Revision 0, Regarding Repair Activities for Reactor Vessel Head Penetration Nozzles and Associated J-Groove Welds
ML16147A3622016-06-17017 June 2016
Relief Request No. 2-TYP-3-RV-04, Revision 0, Regarding Repair Activities for Reactor Vessel Head Penetration Nozzles and Associated J-Groove Welds
ML14363A4092015-01-28028 January 2015
Relief Request No. 1-TYP-4-RV-04 Regarding the Examination Requirements of Code Case N-729-1
ML1202702982012-02-0707 February 2012
Relief Request VRR3 Regarding Solenoid Operated Valve Remote Position Verification Frequency
ML1131304282011-11-22022 November 2011
Relief Request VRR5 Regarding Turbine Driven Auxiliary Feedwater Valve Test Frequency for the 10-Year Inservice Testing Program Interval
ML1126404122011-09-20020 September 2011
Acceptance Review Results for VC Summer Relief Request (ME6879)
ML1107705512011-04-26026 April 2011
Relief Request VRR2 Regarding the 10-Year Inservice Testing Program Interval
ML1104705572011-02-25025 February 2011
Relief Request Regarding an Alternative Weld Repair Method for Reactor Vessel Head Penetrations J-Groove Welds
ML1006807812010-03-12012 March 2010
Third 10-Year ISI Interval Relief Request (ME2608)
L-08-362, Beaver, Units 1 and 2, Proposed Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Paragraph IWA-5244 Examination Requirements2008-12-0202 December 2008
Beaver, Units 1 and 2, Proposed Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Paragraph IWA-5244 Examination Requirements
L-08-207, Proposed Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Inspection Period Extension Requirement2008-09-24024 September 2008
Proposed Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Inspection Period Extension Requirement
L-08-069, Impractical American Society of Mechanical Engineers Code Section XI Weld Examination Requirements (Request Nos. 1-TYP-3-RA-1, 1-TYP-3-RA-2, 1-TYP-3-RA-3)2008-04-0909 April 2008
Impractical American Society of Mechanical Engineers Code Section XI Weld Examination Requirements (Request Nos. 1-TYP-3-RA-1, 1-TYP-3-RA-2, 1-TYP-3-RA-3)
ML0720504882007-09-17017 September 2007
Relief Request No. BV1-PZR-01 Regarding Weld Overlay Repairs on Pressurizer Nozzle Welds
ML0705905552007-04-30030 April 2007
Relief, Relief Request No. BV2-PZR-01 Regarding Weld Overlay Repairs on Pressurizer Nozzle Welds
L-07-056, Requests Approval of Proposed Alternatives and Relief for Inservice Testing Program Ten-Year Update2007-03-28028 March 2007
Requests Approval of Proposed Alternatives and Relief for Inservice Testing Program Ten-Year Update
ML0625801202006-10-0202 October 2006
(BVPS-1 and 2), Inservice Inspection (ISI) Program, Alternative Examination of Reactor Coolant Pipe Welds, Request for Relief No. BV3-RV-2
2024-05-13
[Table view]
Category:Letter
MONTHYEARIR 05000334/20240112024-10-17017 October 2024
License Renewal Phase IV Inspection Report 05000334/2024011
L-24-015, Twenty-Ninth Refueling Outage Inservice Inspection Summary Report2024-09-17017 September 2024
Twenty-Ninth Refueling Outage Inservice Inspection Summary Report
L-24-038, License Amendment Request to Remove the Expired One-Time Action for Valve Leak Repair in Technical Specification (TS) 3.5.2 and to Correct Core Operating Limits Rep Ort (COLR) References2024-09-17017 September 2024
License Amendment Request to Remove the Expired One-Time Action for Valve Leak Repair in Technical Specification (TS) 3.5.2 and to Correct Core Operating Limits Rep Ort (COLR) References
ML24134A1522024-09-17017 September 2024
Exemption from the Requirements of 10 CFR 50.71(e)(4) Final Safety Analysis Report Update Schedule (EPID L-2024-LLE-0005) - Letter
ML24260A1912024-09-16016 September 2024
Operator Licensing Examination Approval
IR 05000334/20240052024-08-29029 August 2024
Updated Inspection Plan for Beaver Valley Power Station, Units 1 and 2 (Reports 05000334/2024005 and 05000412/2024005)
L-24-188, Submittal of Quality Assurance Program Manual, Revision 302024-08-27027 August 2024
Submittal of Quality Assurance Program Manual, Revision 30
L-24-199, Discharge Monitoring Report (NPDES) Permit No. PA00256152024-08-22022 August 2024
Discharge Monitoring Report (NPDES) Permit No. PA0025615
IR 05000334/20244022024-08-22022 August 2024
Security Baseline Inspection Report 05000334/2024402 and 05000412/2024402 (Cover Letter Only)
L-24-186, Response to RAI for Exemption Request from 10 CFR 50.71(e)(4) Final Safety Analysis Update Schedule2024-08-15015 August 2024
Response to RAI for Exemption Request from 10 CFR 50.71(e)(4) Final Safety Analysis Update Schedule
IR 05000334/20240022024-08-0505 August 2024
Integrated Inspection Report 05000334/2024002 and 05000412/2024002
ML24208A0462024-07-26026 July 2024
NRC Office of Investigations Case No. 1-2023-005
L-24-182, Discharge Monitoring Report (NPDES) Permit No. PA00256152024-07-23023 July 2024
Discharge Monitoring Report (NPDES) Permit No. PA0025615
L-24-161, Discharge Monitoring Report (NPDES) Permit No. PA00256152024-07-19019 July 2024
Discharge Monitoring Report (NPDES) Permit No. PA0025615
05000334/LER-2024-004, Reactor Trip Due to Bypass Feedwater Regulating Valve Failure to Control2024-07-17017 July 2024
Reactor Trip Due to Bypass Feedwater Regulating Valve Failure to Control
ML24193A2912024-07-16016 July 2024
Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule
L-24-014, Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling Svstem Evaluation Models2024-07-16016 July 2024
Annual 10 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling Svstem Evaluation Models
IR 05000334/20245012024-07-0808 July 2024
Emergency Preparedness Biennial Exercise Inspection Report 05000334/2024501 and 05000412/2024501
05000334/LER-2024-003, Unanalyzed Containment Bypass Condition Due to Degraded River Water Piping2024-07-0202 July 2024
Unanalyzed Containment Bypass Condition Due to Degraded River Water Piping
L-24-164, BV-2, Post Accident Monitor Report2024-06-27027 June 2024
BV-2, Post Accident Monitor Report
IR 05000334/20244012024-06-26026 June 2024
Material Control and Accounting Program Inspection Report 05000334/2024401 and 05000412/2024401 (Cover Letter Only)
L-24-094, Reactor Vessel Surveillance Capsule Withdrawal Schedule2024-06-24024 June 2024
Reactor Vessel Surveillance Capsule Withdrawal Schedule
L-24-152, Response to NRC Regulatory Issue Summary 2024-01 Preparation and Scheduling of Operator Licensing Examinations2024-06-17017 June 2024
Response to NRC Regulatory Issue Summary 2024-01 Preparation and Scheduling of Operator Licensing Examinations
05000334/LER-2024-002, Automatic Reactor Trip Due to Low Reactor Coolant Flow Caused by Transformer Overexcitation Protection Relay Actuation2024-06-11011 June 2024
Automatic Reactor Trip Due to Low Reactor Coolant Flow Caused by Transformer Overexcitation Protection Relay Actuation
05000334/LER-2024-001, Offsite AC Power Source Inoperable Due to Breaker Failure Resulting in Condition Prohibited by Technical Specification2024-06-0606 June 2024
Offsite AC Power Source Inoperable Due to Breaker Failure Resulting in Condition Prohibited by Technical Specification
ML24135A2282024-05-29029 May 2024
Review of the Spring 2023 Outage Generic Letter 95-05 Voltage-Based Alternate Repair Criteria and Steam Generator F Star Reports
ML24135A1702024-05-29029 May 2024
– Steam Generator Tube Inspection - Review of the Spring 2023 Tube Inspection Reports
L-24-121, Discharge Monitoring Report (NPDES) Permit No. PA00256152024-05-23023 May 2024
Discharge Monitoring Report (NPDES) Permit No. PA0025615
L-24-021, Cycle 30 Core Operating Limits Report2024-05-23023 May 2024
Cycle 30 Core Operating Limits Report
ML24141A1052024-05-20020 May 2024
Senior Reactor and Reactor Operator Initial License Examinations
IR 05000334/20240102024-05-15015 May 2024
Information Request for Quadrennial Baseline Comprehensive Engineering Team Inspection; Notification to Perform Inspection 05000334/2024010 and 05000412/2024010
L-24-107, CFR 50.55a Request 1-TYP-5-RRP-l, for Alternative Method of Nondestructive Examination for River Water Outlet Piping Weld Repair2024-05-13013 May 2024
CFR 50.55a Request 1-TYP-5-RRP-l, for Alternative Method of Nondestructive Examination for River Water Outlet Piping Weld Repair
IR 05000334/20240012024-05-0808 May 2024
Integrated Inspection Report 05000334/2024001 and 05000412/2024001
L-23-269, Application to Revise Technical Specifications to Adopt TSTF-569, Revision of Response Time Testing Definitions2024-05-0707 May 2024
Application to Revise Technical Specifications to Adopt TSTF-569, Revision of Response Time Testing Definitions
L-24-054, Submittal of 2023 Annual Radioactive Effluent Release Report 2023 Annual Radiological Environmental Operating Report and 2023 Annual Environmental Operating Report (Non-Radiological)2024-04-29029 April 2024
Submittal of 2023 Annual Radioactive Effluent Release Report 2023 Annual Radiological Environmental Operating Report and 2023 Annual Environmental Operating Report (Non-Radiological)
L-24-089, Emergency Preparedness Plan2024-04-23023 April 2024
Emergency Preparedness Plan
L-24-088, Discharge Monitoring Report (NPDES) Permit No. PA0025615 for March 20242024-04-22022 April 2024
Discharge Monitoring Report (NPDES) Permit No. PA0025615 for March 2024
ML24101A2752024-04-10010 April 2024
Response to Request for Additional Information Regarding Spring 2023 180-Day Steam Generator Tube Inspection Report
L-24-082, Withdrawal of Exemption Request2024-04-0303 April 2024
Withdrawal of Exemption Request
L-24-013, Annual Notification of Property Insurance Coverage2024-03-26026 March 2024
Annual Notification of Property Insurance Coverage
L-24-064, Submittal of Discharge Monitoring Report (NPDES) Permit No. PA00256152024-03-13013 March 2024
Submittal of Discharge Monitoring Report (NPDES) Permit No. PA0025615
ML24044A0662024-03-0404 March 2024
Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0083 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting
ML24057A0752024-03-0101 March 2024
The Associated Independent Spent Fuel Storage Installations
IR 05000334/20230062024-02-28028 February 2024
Annual Assessment Letter for Beaver Valley Power Station, Units 1 and 2 (Reports 05000334/2023006 and 05000412/2023006)
L-23-264, Request for Exemption from 10 CFR 50.71(e)(4) Final Safety Analysis Report Update Schedule2024-02-23023 February 2024
Request for Exemption from 10 CFR 50.71(e)(4) Final Safety Analysis Report Update Schedule
CP-202300502, Notice of Planned Closing of Transaction and Provision of Documents to Satisfy Order Conditions2024-02-23023 February 2024
Notice of Planned Closing of Transaction and Provision of Documents to Satisfy Order Conditions
L-24-050, Retrospective Premium Guarantee2024-02-22022 February 2024
Retrospective Premium Guarantee
05000412/LER-2023-003-01, Missile Barrier Door Left Open Resulting in a Loss of Safety Function2024-02-20020 February 2024
Missile Barrier Door Left Open Resulting in a Loss of Safety Function
IR 05000334/20230042024-02-12012 February 2024
Integrated Inspection Report 05000334/2023004 and 05000412/2023004
ML24025A0922024-01-25025 January 2024
Requalification Program Inspection
2024-09-17
[Table view]
Category:Safety Evaluation
MONTHYEARML24193A2912024-07-16016 July 2024
Revision to Reactor Vessel Surveillance Capsule Withdrawal Schedule
ML23198A3592023-10-0202 October 2023
Issuance of Amendment Nos. 322 and 212 Analytical Methodology to the Core Operating Limits Report for a Full Spectrum Loss of Coolant Accident (EPID L-2022-LLA-0129) - Nonproprietary
ML23237B4282023-09-28028 September 2023
Energy Harbor Nuclear Corp. - Vistra Operations Company LLC - Enclosure 2, Draft Conforming License Amendments
ML23237B4302023-09-28028 September 2023
Energy Harbor Nuclear Corp. - Vistra Operations Company LLC - Enclosure 4, Safety Evaluation for Transfer of Licenses and Draft Conforming License Amendments (EPID L-2023-LLM-0000) (Public)
ML23249A1842023-09-18018 September 2023
Authorization and Safety Evaluation for Alternative Request No. 2-TYP-4-RV-06
ML23188A0982023-07-17017 July 2023
Correction to the Safety Evaluation Issued Related to Amendment Nos. 321 and 211 Consolidate Fuel Decay Time Technical Specifications in a New Limiting Condition for Operation Titled Decay Time
ML23102A1472023-05-22022 May 2023
Issuance of Amendment Nos. 321 and 211 Consolidate Fuel Decay Time Technical Specifications in a New Limiting Condition for Operation Titled Decay Time
ML23019A0032023-03-16016 March 2023
Issuance of Amendment Nos. 320 and 210 Adoption of Technical Specifications Task Force (Tstf) Traveler Tstf-501, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control
ML23062A5212023-03-0606 March 2023
Issuance of Amendment No. 319 Revise Technical Specification (TS) 3.5.2, ECCS Operating, Limiting Condition for Operation (LCO) 3.5.2,
ML22277A8142022-10-0707 October 2022
Correction of Errors in Safety Evaluation Associated with License Amendment Nos. 315 and 205 Regarding Changes to the Emergency Preparedness Plan
ML22210A0102022-09-16016 September 2022
Energy Harbor Fleet- Issuance of Amendments Regarding Adoption of TSTF 554, Revise Reactor Coolant Leakage Requirements
ML22235A7672022-09-0101 September 2022
Correction of Errors in Safety Evaluation Associated with License Amendment Nos. 315 and 205 Regarding Changes to the Emergency Preparedness Plan
ML22222A0862022-09-0101 September 2022
Issuance of Amendment Nos. 317 and 208 Revise Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start and Bus Separation Instrumentation
ML22202A4642022-06-29029 June 2022
Emergency Plan Safety Evaluation
ML22140A2092022-06-28028 June 2022
Issuance of Amendment No. 207 Correct TS 3.1.7 Change Made by TSTF-547
ML22095A2352022-05-10010 May 2022
Issuance of Amendment Nos. 316 and 206 Revise Technical Specification 5.6.3, Core Operating Limits Report (COLR)
ML21286A7822022-05-0606 May 2022
Issuance of Amendment Nos. 315 and 205 Regarding Changes to the Emergency Plan
ML22077A1342022-05-0202 May 2022
Issuance of Amendment Nos. 314 and 204 Revise Technical Specifications to Adopt TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections
ML22082A2532022-03-29029 March 2022
Correction Letter - Issuance of Authorization of Proposed Alternative to Use ASME OM Code Case OMN-27
ML22012A2972022-01-21021 January 2022
Proposed Alternative to Use ASME OM Code Case OMN-27
ML21197A0092021-11-0101 November 2021
Issuance of Amendment Nos. 313 and 203 Reactor Coolant System, Pressure and Temperature Limits Report
ML21214A2752021-10-15015 October 2021
Issuance of Amendment Nos. 312 and 202 Atmospheric Dump Valves
ML21153A1762021-06-30030 June 2021
Issuance of Amendment No. 201 Revision of Technical Specifications Related to Steam Generator Tube Inspection, and Repair Methods
ML21075A1132021-04-16016 April 2021
Issuance of Amendment Nos. 311, 200, and 302 to Incorporate the Applicable Standard Technical Specification 5.2.2, Unit Staff, Into the Technical Specifications
ML21070A0002021-03-22022 March 2021
Issuance of Amendment 310, Revise Technical Specification 5.5.5.1, Unit 1 SG Program, to Defer Spring 2021 Refueling Outage Steam Generator Inspections to Fall 2022 Refueling Outage
ML20346A0222021-03-10010 March 2021
Issuance of Amendment Nos. 309 and 199 to Change Technical Specifications to Implement New Surveillance Methods for the Heat Flux Hot Channel Factor
ML20335A0522021-02-18018 February 2021
Issuance of Amendment Nos. 308 and 198 to Modify Certified Fuel Handler Related Technical Specifications for Permanently Defueled Condition
ML20345A2362021-01-28028 January 2021
Issuance of Amendment Nos. 307 and 197 to Add Containment Sump Technical Specifications to Address GSI-191 Issues
ML20335A0232020-12-28028 December 2020
Issuance of Amendment Nos. 306 and 196 to Remove License Conditions B and C Related to the Irradiated Fuel Management Plans
ML20285A2662020-10-21021 October 2020
Correction to Safety Evaluation for Amendment Nos. 305 and 195 Issued September 23, 2020, Modify Primary and Secondary Coolant Activity Technical Specifications
ML20279A4402020-10-0808 October 2020
Energy Harbor Fleet-Beaver Valley Power Station; Davis-Besse Nuclear Power Station, and Perry Nuclear Power Plant - Request to Use a Provision of a Later Edition of the ASME Engineers Boiler and Pressure Vessel Code, Section XI
ML20213A7312020-09-23023 September 2020
Issuance of Amendment Nos. 305 and 195 to Modify Primary and Secondary Coolant Activity Technical Specifications
ML20223A0142020-08-20020 August 2020
Issuance of Relief Request SSR-1, Revision 0, from the Requirements of the ASME Code (EPID L-2020-LLR-0050 (COVID-19))
ML20206K8652020-08-0707 August 2020
Relief Requests 2-TYP-4-RV-06 and 2-TYP-4-RV-07 from the Requirements of the ASME Code (EPID L-2020-LLR-0053 and EPID L-2020-LLR-0054 (COVID-19))
ML20153A0142020-07-16016 July 2020
Relief from the Requirements of the American Society of Mechanical Engineers Code Regarding Request VRR4 (EPID L-2020-LLR-0052 (COVID-19))
ML20145A0002020-06-0303 June 2020
Relief from the Requirements of the ASME Code for Requests VRR6 and VRR5 (EPID L-2020-LLR-0049 and EPID L-2020-LLR-0051 (COVID-19))
ML20080J7892020-04-28028 April 2020
Relief Requests 2 TYP-3-B3.110-1, 2-TYP-3-C2.21-1, 2-TYP-3-C1.30-1, and 2-TYP-3-RA-1 Regarding Weld Examination Coverage for the Third Inservice Inspection Interval
ML20079F8162020-03-26026 March 2020
Relief Requests 1-TYP-4-C2.21-1 and 1-TYP-4-RA-1 Regarding Weld Examination Coverage for the Fourth Inservice Inspection Interval
ML19336A0282019-12-18018 December 2019
Safety Evaluation Irradiated Fuel Management Plans
ML19305B1312019-12-0202 December 2019
Firstenergy Nuclear Operating Company - Enclosure 6, Non-Proprietary Safety Evaluation, Direct and Indirect Transfer of Licenses and Draft Conforming Amendments for Beaver Valley Units 1 and 2, Davis-Besse Unit 1, and Perry Unit 1
ML19275E2942019-10-16016 October 2019
Issuance of Relief Request Proposed Alternative to Reactor Vessel Nozzle Weld Examination Frequency Requirements in Lieu of Specific ASME Code Requirements
ML19028A0302019-04-11011 April 2019
FENOC - Beaver Valley Power Station, Unit 1 & 2; Davis-Besse Nuclear Power Station Unit 1; and Perry Nuclear Power Plant Unit 1 - Approval of Certified Fuel Handler Training and Retraining Program
ML18348B2062019-02-25025 February 2019
Issuance of Amendment No. 193 Revise Steam Generator Technical Specifications
ML18249A1692018-09-0707 September 2018
Seismic Hazard Mitigation Strategies Assessment (CAC Nos. MF7800 and MF7801; EPID L-2016-JLD-0006)
ML18227A7332018-08-27027 August 2018
Request for Relief from the Requirements of the ASME Code
ML18205A2922018-08-10010 August 2018
Correction of Errors in Safety Evaluation Associated with Relief Request 2-TYP-4-RV-02
ML18179A4672018-07-30030 July 2018
FENOC-Beaver Valley Power Station, Unit No. 1 and 2; Davis-Besse Nuclear Power Station, Unit 1, and Perry Nuclear Power Plant, Unit 1 - Issuance of Amendments Request to Adopt TSTF-529, Clarify Use and Application Rules
ML18178A1122018-07-0202 July 2018
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 2, 2017 Mr. Marty L. Richey, Site Vice President FirstEnergy Nuclear Operating Company Beaver Valley Power Station Mail Stop A-BV-SEB1 P.O. Box 4, Route 168 Shippingport, PA 15077
SUBJECT:
BEAVER VALLEY POWER STATION, UNIT NO. 1 -RELIEF FROM THE REQUIREMENTS OF THE ASME CODE (CAC NO. MF8531)
Dear Mr. Richey:
By letter dated October 24, 2016, as supplemented by letter dated January 13, 2017 (Agencywide Documents Access and Management System Accession Nos. ML16298A289 and ML17013A483, respectively), FirstEnergy Nuclear Operating Company (FENOC or the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to certain American Society of Mechanical Engineers (ASME) Code for Operating and Maintenance of Nuclear Power Plants (OM Code) requirements at Beaver Valley Power Station, Unit No. 1.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee requested to use the proposed alternative on the basis that complying with the specified requirement would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. The proposed alternative is to delay a relief valve test to the maintenance and refueling outage of the spring of 2018.
The NRC staff has reviewed the subject request and determined, as set forth in the enclosed safety evaluation, that FENOC has adequately addressed all of the regulatory requirements set for in 10 CFR 50.55a and that the proposed alternative provides reasonable assurance that the affected component is operationally ready. Accordingly, the NRC staff concludes that FENOC has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2) and is in compliance with ASME OM Code requirements. Therefore, the NRC staff authorizes the proposed alternative request to test valve RV-1 RH-721 during the next refueling outage, which is currently scheduled for the spring of 2018.
All other ASME OM Code requirements for which relief was not specifically requested and approved in this relief request remain applicable.
M. Richey If you have any questions, please contact the Project Manager, Taylor Lamb, at 301-415-7128 or Taylor.Lamb@nrc.gov.
Sincerely,
~j)(A Stephen S. Koenick, Acting Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-334
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv
REQUEST FOR ALTERNATIVE VRR6 REGARDING RESIDUAL HEAT REMOVAL SYSTEM VALVE TESTING FIRSTENERGY NUCLEAR OPERATING COMPANY BEAVER VALLEY POWER STATION. UNIT NO. 1 DOCKET NO. 50-334
1.0 INTRODUCTION
By letter dated October 24, 2016, as supplemented by letter dated January 13, 2017 (Agencywide Documents Access and Management System Accession Nos. ML16298A289 and ML17013A483, respectively), FirstEnergy Nuclear Operating Company (FE NOC or the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC or the Commission) for the use of alternatives to certain American Society of Mechanical Engineers (ASME) Code for Operating and Maintenance of Nuclear Power Plants (OM Code) requirements at Beaver Valley Power Station, Unit No. 1 (BVPS-1).
Specifically, pursuant to Title 1O of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee requested to use the proposed alternative on the basis that complying with the specified requirement would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. The proposed alternative is to delay a relief valve test to the maintenance and refueling outage of the spring of 2018.
2.0 REGULATORY EVALUATION
The regulations in 10 CFR 50.55a(f), "lnservice testing requirements," require, in part, that inservice testing (IST) of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda, except where alternatives have been authorized pursuant to 10 CFR 50.55a(z)(1) or 10 CFR 50.55a(z)(2).
In proposing alternatives, a licensee must demonstrate that the proposed alternatives provide an acceptable level of quality and safety (10 CFR 50.55a(z)(1)), or compliance would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety
( 10 CFR 50.55a(z)(2) ).
Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request, and the Commission to authorize, the alternative requested by the licensee.
Enclosure
3.0 TECHNICAL EVALUATION
3.1 Licensee Relief Request No. VRR6 Alternative testing is requested for the BVPS-1 residual heat removal (RHR) pump relief valve RV-1 RH-721 (Class 2, Category C).
Applicable ASME OM Code requirements, as stated by the licensee, are:
Mandatory Appendix I, 1-1350, "Test Frequency, Class 2 and 3 Pressure Relief Valves," Paragraph (c), "Requirements for Testing Additional Valves," states in part:
Additional valves shall be tested in accordance with the following requirements:
- 1. For each valve tested for which as-found set-pressure (first test actuation) exceeds the greater of either the +/- tolerance limit of the Owner established set-pressure acceptance criteria of 1-131 O(e) or +/-3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group.
The licensee states:
Reason for Request
Charging system letdown relief valve RV-1 CH-203 has a cold differential test pressure (CDTP) of 606 pounds per square inch gauge (psig) with an upper acceptable limit of plus 3 percent or 624 psig. The CDTP is a temperature compensated test pressure used to account for the difference between the ambient temperature at the test stand and the higher temperature at the tested valve's installed location in the plant. During the 24th Beaver Valley Power Station, Unit No. 1 (BVPS-1 ), maintenance and refueling outage (1 R24) when RV-1 CH-203 was tested on the test stand at ambient conditions, the relief valve lifted at 631.7 psig, which was 4.25 percent above its CDTP. This exceeded the plus 3 percent limit specified in Paragraph l-1350(c)(1) requiring the only other valve in the sample group, RV-1 RH-721, to be tested. However, because the ASME OM Code, Appendix I, Paragraph l-1350(c)(1) allows for the owner to establish set-pressure acceptance criteria (a plus or minus tolerance limit), a limit of plus 5 percent was calculated after the valve exceeded the plus 3 percent limit specified in the ASME OM Code. Relief valve RV-1 CH-203 test results were within the plus 5 percent acceptance criteria. Therefore, valve RV-1 RH-721 was not tested. When it was determined that FENOC was not permitted to provide an owner specified limit after the valve was tested, the plant lineup was being established for plant startup with the reactor core fully loaded with fuel.
Testing of RV-1 RH-721 cannot be performed in-place. The relief valve must be removed from the system and tested on a test stand. In order to remove and test RV-1 RH-721, the RHR system would have to be shut down and the entire
system drained. This is normally performed when the RHR system is not required to be in operation, which is when fuel is removed from the reactor core.
Based on the plant lineup and the need to remove the valve from the system to perform the test, the sample size could not be expanded as required and is not in compliance with the ASME OM Code. Therefore, a delay is proposed to test RV-1 RH-721 during the 25th BVPS-1 maintenance and refueling outage (1 R25) when fuel is removed from the core (1 R25 is scheduled for the spring of 2018).
To unload the reactor core in order to test RV-1 RH-721 provides a hardship without a compensating increase in quality and safety.
Proposed Alternative and Basis for Use The proposed alternative is to delay testing of relief valve RV-1 RH-721 until 1R25. At that time, the RHR system can be drained, and RV-1 RH-721 can be tested.
Valve RV-1 RH-721 is considered operationally ready until tested during 1R25 for the following reasons:
- 1) A work order history review shows that two different serial-numbered (S/N) valves have been swapped in and out of the installed location for RV-1-RH-721 as follows:
1R 11 (April 1996) S/N 53237M 1 installed (set at 610 psig) 1R13 (March 2000) S/N 53237M1 removed (as-found set at 616 psig),
and S/N N69973-01-0001 installed (set at 608 psig) 1R15 (March 2003) S/N N69973-01-0001 removed (as-found set at 51 O psig due to leakage), and S/N 53237M1 installed (set at 610 psig) 1R17 (January 2006) S/N 53237M1 removed (as-found set at 603 psig),
and SIN N69973-01-0001 installed (set at 610 psig) 1R20 (October 2010) S/N N69973-01-0001 removed (as-found set at 611.2 psig), and S/N 53237M1 installed (set at 606 psig)
The setpoint testing listed above was performed on a test stand at ambient conditions.
The preceding data shows that valve S/N N69973-01-0001 has drifted high approximately 2 psig over a 3 to 4 year period while valve S/N 53237M1 has drifted high approximately 6 psig over a 4 year period. If the approximate 6 psig
high drift for valve S/N 53237M1 is extrapolated over a 7.5 year period from 1R20 to 1R25, RV-1RH-721 would be expected to lift no higherthan 618 psig.
Therefore, based on past performance, there is reasonable assurance that RV-1 RH-721 would not lift greater than plus 3 percent above its set-pressure when tested during the next maintenance and refueling outage.
- 2) Other than a three-year period from 1R13 to 1R15 when the valve was found to lift low due to leakage, RV-1RH-721 has shown a history of consistent performance. Therefore, there is reasonable assurance that valve RV-1 RH-721 will continue to be operationally ready until the next scheduled test during 1R25. Further, the test interval from 1R20 to 1R25 (7.5 years) is conservatively shorter than the maximum 10-year test interval requirement of ASME OM Code, Mandatory Appendix I, Paragraph l-1350(a).
In addition, during operation at power, a surveillance verification log monitors annunciator window A 1-125, "Residual Heat Removal Pump Discharge Pressure HIGH." This. annunciator alarms if pressure reaches 550 psig and provides corrective actions to take in accordance with an alarm response procedure. The corrective actions would relieve pressure before it reaches the RV-1 RH-721 set-pressure of 600 psig.
The proposed alternative to ASME OM Code, Mandatory Appendix I, paragraph l-1350(c) would delay the test of relief valve RV-1 RH-721 until 1R25.
Therefore, there is reasonable assurance that relief valve RV-1 RH-721 will continue to be operationally ready until 1R25.
The proposed alternative is requested for use during the remainder of the fourth 10-year IST interval.
3.2 NRC Staff Evaluation ASME OM Code, 2001 Edition through 2003 Addenda, Section Mandatory Appendix I, "lnservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants,"
paragraph l-1350(a), 10-year test Interval states that:
Class 2 and 3 pressure relief valves, with the exception of PWR
[pressurized-water reactor] main steam safety valves, shall be tested every 10 years, starting with initial electric power generation. No maximum limit is specified for the number of valves to be tested during any single plant operating cycle; however, a minimum of 20% of the valves from each valve group shall be tested within any 48-month interval. This 20% shall consist of valves that have not been tested during the current 10-year test interval, if they exist.
Requirements for testing additional valves in a group is detailed in paragraph l-1350(c)(1),
which states that, "For each valve tested for which the as-found set-pressure (first test actuation) exceeds the greater of either the+/- tolerance limit of the Owner established set-pressure acceptance criteria of l-1310(e) or +/-3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group."
As noted in the alternative request, relief valve RV-1 CH-203 was recently tested pursuant to the required ASME OM Code test interval and failed to meet the standard ASME OM Code tolerance acceptance value of +/-3 percent of valve nameplate set-pressure. The standard ASME OM Code tolerance acceptance value was used because the licensee did not have an owner-established set-pressure acceptance criteria specified. The valve failure now required additional valves of the valve group to be tested. A valve group is defined in ASME OM Code, 2001 Edition through 2003 Addenda, as "Valves of the same manufacturer, type, system application, and service media." Valve RV-1 CH-203 has only one other valve associated with it in its group, which is RV-1 RH-721.
Relief valve RV-1 RH-721 protects the RHR system. The valve cannot be tested in situ and must be removed from the system. To remove the valve from the system will require draining the RHR system, which is normally performed when the fuel is removed from the reactor core.
The next scheduled fuel removal is the spring of 2018. To perform the required ASME OM Code test now would represent a hardship, without a compensating increase in quality and safety.
The licensee proposes to delay testing RV-1 RH-721 until the next refuel outage when the RHR system can be removed from service after the reactor core is off-loaded. The NRC staff reviewed the performance history of the setpoint testing. Considering the sample, in which there was no instrument drift high above 1 percent, the NRC staff has reasonable assurance that the valve would not lift greater than +3 percent. Furthermore, the NRC staff credits the fact that at power, a surveillance verification log monitors annunciator window A 1-125 and that the annunciator alarms if pressure reaches 550 psig. The resulting corrective actions would relieve pressure before they reach the RV-1 RH-721 set-pressure of 600 psig. Therefore, the NRC staff concludes that the proposed alternative is acceptable.
4.0 CONCLUSION
As set forth above, the NRC staff determined that the proposed alternative provides reasonable assurance that the affected component is operationally ready. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(z)(2).
All other ASME OM Code requirements for which relief was not specifically requested and approved in the subject requests for relief remain applicable.
Therefore, the NRC staff authorizes the BVPS-1 proposed alternative request to test valve RV-1 RH-721 during the next refueling outage, which is currently scheduled for the spring of 2018.
Principal Contributor: M. Farnan Date: March 2, 2017
ML17047A610 *by e-mail dated OFFICE DORL/LP LI/PM DORL/LPLl/LA DE/EPNB/BC* DORL/LP LI/BC DORL/LP LI/PM NAME TLamb LRonewicz DAiiey SKoenick TLamb DATE 02/21/2017 02/21/2017 02/13/2017 02/28/2017 03/02/2017