L-20-060, CFR 50.55a Request Number: VRR4, Revision 0, Containment Isolation Valve Test Frequency
| ML20093C288 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/02/2020 |
| From: | Penfield R Energy Harbor Nuclear Corp |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-20-060 | |
| Download: ML20093C288 (8) | |
Text
Energy Harbor Nuclear Corp.
Beaver Valley Power Station P. O. Box 4 Shippingport, PA 15077 10 CFR 50.55a April 2, 2020 L-20-060 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Beaver Valley Power Station, Units No. 1 and 2 Docket No. 50-334, License No. DPR-66 Docket No. 50-412, License No. NPF-73 10 CFR 50.55a Request Number: VRR4, Revision 0, Containment Isolation Valve Test Frequency In accordance with 10 CFR 50.55a(z)(2), Energy Harbor Nuclear Corp. hereby requests Nuclear Regulatory Commission (NRC) approval of request VRR4, Revision 0, that proposes an alternative test frequency for certain valves in the Beaver Valley Power Station, Unit No. 1 (BVPS-1) and Unit No. 2 (BVPS-2) Inservice Testing Programs for Pumps and Valves (IST Programs).
As a result of the hardship produced by the recent pandemic and the resulting national state of emergency, Energy Harbor Nuclear Corp. is requesting expedited approval of VRR4 for BVPS-1 and BVPS-2. The proposed alternative would be implemented during the fifth 10-year inservice test interval at BVPS-1 and the fourth 10-year inservice test interval at BVPS-2.
To support the startup and critical generation of BVPS-2 from its scheduled refuel outage, Energy Harbor Nuclear Corp. requests approval of the proposed alternative by April 12, 2020.
The enclosed request identifies the affected components, applicable code requirements, and a description and basis for the proposed alternative.
Rod L. Penfield Site Vice President, Beaver Valley Nuclear 724-682-5234
Beaver Valley Power Station, Unit Nos. 1 and 2 L-20-060 Page 2 There are no regulatory commitments contained in this submittal. If there are any questions or additional information is required, please contact Mr. Phil H. Lashley, Acting Manager - Nuclear Licensing and Regulatory Affairs, at (330) 315-6808.
Sincerely, Rod L. Penfield
Enclosure:
Beaver Valley Power Station, Unit Nos. 1 and 2, 10 CFR 50.55a Request Number VRR4, Revision 0 cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative
Beaver Valley Power Station, Unit Nos. 1 and 2 10 CFR 50.55a Request Number: VRR4, Revision 0 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2)
Page 1 of 6
-- Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety --
- 1. ASME Code Components Affected Table 1 Valve ID Valve Description Class Category 1SI-83 Hot Legs HHSI Supply Check 1
A/C MOV-1SI-869A HHSI to RCS Hot Legs Isolation 1
A 1CH-31 Regen H/X Inlet Check 2
A/C MOV-1CH-289 Chg PP Disch Hdr to Regen H/X 2
A 1SI-84 Hot Legs HHSI Supply Check 1
A/C MOV-1SI-869B HHSI to RCS Hot Legs Isolation 1
A 1CH-181 RCP 1A Seal Supply Check 2
A/C MOV-1CH-308A RCP 1A Seal Injection Isolation 2
A 1CH-182 RCP 1B Seal Supply Check 2
A/C MOV-1CH-308B RCP 1B Seal Injection Isolation 2
A 1CH-183 RCP 1C Seal Supply Check 2
A/C MOV-1CH-308C RCP 1C Seal Injection Isolation 2
A/C FCV-1CH-160 Fill Header Flow Control Valve 1
A MOV-1SI-890A A LHSI to RCS Hot Legs Isolation 1
A MOV-1SI-890C LHSI to RCS Cold Legs Isolation 1
A MOV-1SI-890B B LHSI to RCS Hot Legs Isolation 1
A MOV-1SI-860A A LHSI PP CNMT Sump Suct Isol 2
A MOV-1SI-860B B LHSI PP CNMT Sump Suct Isol 2
A 1SI-95 Cold Legs HHSI Supply Check 1
A/C MOV-1SI-836 HHSI to RCS Cold Legs Isolation 1
A 1SI-94 Cold Legs BIT/HHSI Supply Chk 1
A/C MOV-1SI-867C Boron Injection Tank Outlet Isol 1
A MOV-1SI-867D Boron Injection Tank Outlet Isol 1
A 2SIS*83 HHSI Check Vlv to RCS Hot Legs 2
A/C 2SIS*MOV869A HHSI Hot Legs Injection Isolation 2
A 2CHS*31 Charging Header Isolation Check 2
A/C 2CHS*MOV289 Normal Charging Header Isolation 2
A 2SIS*84 HHSI Check Vlv to RCS Hot Legs 2
A/C 2SIS*MOV869B HHSI Hot Legs Injection Isolation 2
A 2SIS*94 HHSI Chk Vlv to RCS Cold Legs 2
A/C 2SIS*MOV836 High Head to Cold Legs Inj Isol 2
A 2SIS*MOV840 High Head to Cold Legs Inj Isol 2
A 2CHS*474 RCP 21A Seal Supply CNMT Chk 2
A/C 2CHS*MOV308A RCP 21A Seal Water Injection Isol 2
A 2CHS*476 RCP 21B Seal Supply CNMT Chk 2
A/C
Beaver Valley Power Station, Unit Nos. 1 and 2 10 CFR 50.55a Request Number: VRR4, Revision 0 Page 2 of 6 Table 1 Valve ID Valve Description Class Class 2CHS*MOV308B RCP 21B Seal Water Injection Isol 2
A 2CHS*475 RCP 21C Seal Supply CNMT Chk 2
A/C 2CHS*MOV308C RCP 21C Seal Water Injection Isol 2
A 2CHS*472 Loop Fill CNMT Isolation Check 2
A/C 2CHS*FCV160 RCS Loop Fill Hdr Flow Cont Vlv 2
A 2SIS*95 HHSI Chk Vlv to RCS Cold Legs 2
A/C 2SIS*MOV867C HHSI PP Isol to Cold Leg Injection 2
A 2SIS*MOV867D HHSI PP Isol to Cold Leg Injection 2
A Table 2 Valve ID Valve Description Class Category MOV-1SI-885A LHSI PP 1A Min Flow Line (Train A) 2 A
MOV-1SI-885B LHSI PP 1B Min Flow Line (Train A) 2 A
MOV-1SI-885C LHSI PP 1B Min Flow Line (Train B) 2 A
MOV-1SI-885D LHSI PP 1A Min Flow Line (Train B) 2 A
1SI-27 Charging Pump RWST Supply Check 2
A/C MOV-1CH-115B RWST Outlet to Chg PP Suction Hdr Isol 2
A MOV-1CH-115D RWST Outlet to Chg PP Suction Hdr Isol 2
A 2SIS*6 LHSI Pump 21A Discharge Check 2
A/C 2SIS*MOV8809A LHSI Pump 21A Suction Isolation 2
A 2SIS*MOV8890A LHSI Pump 21A Min Flow Recirc Isolation 2
A 2SIS*7 LHSI Pump 21B Discharge Check 2
A/C 2SIS*MOV8809B LHSI Pump 21B Suction Isolation 2
A 2SIS*MOV8890B LHSI Pump 21B Min Flow Recirc Isolation 2
A 2SIS*27 Check Valve to HHSI Pumps from RWST 2
A/C 2CHS*LCV115B Charging Pump Suction from RSWT 2
A 2CHS*LCV115D Charging Pump Suction from RSWT 2
A Table 3 Valve ID Valve Description Class Category 1CCR-289 RCP 1A Thermal Barrier Cooler Inlet Chk 3
A/C TV-1CC-107A RCP 1A Thermal Barrier Cooler Outlet Isol 3
A 1CCR-290 RCP 1B Thermal Barrier Cooler Inlet Chk 3
A/C TV-1CC-107B RCP 1B Thermal Barrier Cooler Outlet Isol 3
A 1CCR-291 RCP 1C Thermal Barrier Cooler Inlet Chk 3
A/C TV-1CC-107C RCP 1C Thermal Barrier Cooler Outlet Isol 3
A 2CCP*289 RCP 21A Thermal Barrier Clr Supply Chk 3
A/C 2CCP*MOV107A RCP 21A Thermal Barrier Clr Disch Isol 3
A 2CCP*290 RCP 21B Thermal Barrier Clr Supply Chk 3
A/C 2CCP*MOV107B RCP 21B Thermal Barrier Clr Disch Isol 3
A 2CCP*291 RCP 21C Thermal Barrier Clr Supply Chk 3
A/C 2CCP*MOV107C RCP 21C Thermal Barrier Clr Disch Isol 3
A Valve identification (ID) numbers that start with the number 2 are BVPS-2 valves, other listed valve ID numbers are BVPS-1 valves.
Beaver Valley Power Station, Unit Nos. 1 and 2 10 CFR 50.55a Request Number: VRR4, Revision 0 Page 3 of 6 Acronyms used in Tables 1, 2, and 3 BIT Boron Injection Tank Chg Charging Chk Check CHS Charging System Clr Cooler CNMT Containment Cont Control Disch Discharge FCV Flow Control Valve Hdr Header H/X Heat Exchanger HHSI High Head Safety Injection Inj Injection Isol Isolation LHSI Low Head Safety Injection Min Minimum MOV Motor operated valve PP Pump RCP Reactor Coolant Pump RCS Reactor Coolant System Recirc Recirculation Regen Regenerative RWST Refueling Water Storage Tank SI Safety Injection SIS Safety Injection System Suct Suction Vlv Valve
2. Applicable Code Edition and Addenda
American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM)
Code-2004 Edition, with Addenda through OMb-2006.
3. Applicable Code Requirement
ISTC-3630, Leakage Rate for Other Than Containment Isolation Valves (a) Frequency, states that tests shall be conducted at least once every 2 years.
4. Reason for Request
The valves in Table 1 above are containment isolation valves (CIVs) that are expected to remain water filled following a loss of coolant accident (LOCA). Therefore, they are not subject to Type-C leakage tests as noted in the Beaver Valley Power Station, Unit No. 1 (BVPS-1) and Unit No. 2 (BVPS-2) Licensing Requirements Manuals, Table 3.6.1-1, Containment Penetrations, but instead are leakage tested once every two years in accordance with Paragraph ISTC-3630(a) of the ASME OM Code using water.
The valves in Table 2 above are valves that are being leakage tested in response to Information Notice 91-56, Potential Radioactive Leakage to Tank Vented to Atmosphere. These valves have been determined to have a safety function with
Beaver Valley Power Station, Unit Nos. 1 and 2 10 CFR 50.55a Request Number: VRR4, Revision 0 Page 4 of 6 respect to control room and offsite dose limits and will prevent leakage of containment sump water from the emergency core cooling system (ECCS) recirculation lines to the refueling water storage tank (RWST) which is vented to atmosphere. These non-CIVs are currently leakage tested once every two years in accordance with Paragraph ISTC-3630(a) of the ASME OM Code using water.
The valves in Table 3 above are reactor coolant pump thermal barrier cooler isolation valves that isolate the lower pressure component cooling water system from the higher pressure reactor coolant system in the event of a thermal barrier cooler rupture. These non-CIVs are currently leakage tested once every two years in accordance with Paragraph ISTC-3630(a) of the ASME OM Code using water.
The valves identified in Tables 1, 2 and 3 above are all non-Type-C tested valves in water applications that are leakage tested using water at least once every two years.
Although they are not specifically included in the scope for performance based testing as described in NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50 Appendix J, and are not leakage tested in accordance with 10 CFR 50 Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Option B, Performance Based Requirements, Energy Harbor Nuclear Corp. intends to implement an approach similar to the performance based testing frequency of Option B instead of performing leakage testing of these valves once every two years at BVPS-1 and BVPS-2.
The reason for requesting this relief is dose reduction to comport with NRC and industry as low as reasonably achievable (ALARA) radiation dose principles.
Additionally, the extended test frequency will reduce the amount of out of service time for certain high head and low head safety injection valves in the boration flow paths during a refueling outage.
On March 13, 2020, the President of the United States declared a national emergency due to the spread and infectious nature of the COVID-19 virus and resulting pandemic.
The most recent guidance from the Centers for Disease Control and Prevention (CDC) includes recommendations for social distancing by maintaining approximately six feet from other personnel to limit the spread of the virus. On March 28, 2020, the Governor of Pennsylvania issued a Stay at Home order for Beaver County and the surrounding counties of Allegheny and Butler.
To prevent the spread of COVID-19 at BVPS, Energy Harbor Nuclear Corp. intends to reduce the amount of personnel on-site, which will pose a hardship for completing the currently planned 2R21 refueling outage work scope. Energy Harbor Nuclear Corp. is also contingency planning in case some of its workforce becomes unavailable due to the COVID-19 outbreak. With the current work scope and potential loss of personnel, there is the potential that the company may not be able to complete the refueling outage in a timely manner, which could negatively impact critical infrastructure that is needed during this time.
Beaver Valley Power Station, Unit Nos. 1 and 2 10 CFR 50.55a Request Number: VRR4, Revision 0 Page 5 of 6
5. Proposed Alternative and Basis for Use
The proposed alternative is to perform leakage testing of the valves identified in Tables 1, 2 and 3 at intervals based on valve performance ranging from every refueling outage to every fourth refueling outage. The specific interval for each valve would be a function of its performance and would be established in a manner consistent with the CIV process under 10 CFR 50 Appendix J, Option B. Valves that have demonstrated good performance for two consecutive cycles may have their test interval extended to every third refueling, not to exceed 60 months. Valves that have demonstrated good performance for three consecutive cycles may have their test interval extended to every fourth refueling, not to exceed 75 months, if additional considerations as specified in NEI 94-01, Revision 3-A, Section 11.3.2, Programmatic Controls, are met. Any leakage test failure would require the valve to be returned to the initial test frequency of every refueling outage until good performance can again be re-established.
The valves listed in Table 1 and Table 2 are arranged with two valves in series. This valve arrangement ensures that leakage outside of containment or leakage from the containment sump to the refueling water storage tank that is vented to atmosphere will be kept to a minimum. One exception is for BVPS-1 containment penetration numbers 68 and 69 that are located between the water filled containment sump (that is, post LOCA) and one valve (that is, MOV-ISI-860A, and MOV-ISI-860B, respectively) just outside the containment penetrations. A second exception is the RCP thermal barrier cooler valves because only one Category A valve is located on either side of each thermal barrier cooler.
As stated on page 22 of NEI 94-01, Revision 3-A, NUREG-1493, "Performance-Based Containment Leak-Test Program, provided the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate testing intervals. NUREG-1493 found the effect of Type C testing on overall accident risk is small and concluded that performance-based alternatives to local leakage rate testing requirements are feasible without significant risk impacts; and although extended testing intervals led to minor increases in potential off-site dose consequences, the actual decrease in on-site (worker) doses exceeded (by at least an order of magnitude) the potential off-site dose increases.
Option B of Appendix J to 10 CFR 50 provides performance-based primary reactor containment leakage-rate test requirements. NEI 94-01, Revision 3-A, allows for an extended leak test interval of up to 75 months. Although NEI 94-01 does not address seat leakage testing with water, it has been determined that extending the frequency for leak rate testing provides a low level of risk, therefore, the proposed alternative method would provide an acceptable level of quality and safety.
A review of recent historical data identified that leak testing each of the valves listed in Section 1 every refueling outage results in a total personnel dose of approximately 75
Beaver Valley Power Station, Unit Nos. 1 and 2 10 CFR 50.55a Request Number: VRR4, Revision 0 Page 6 of 6 millirem at BVPS-1 and 96 millirem at BVPS-2. The proposed extended test interval (assuming all valves are on an extended frequency) would provide a total dose savings at the site of approximately 0.5 Rem over the course of four refueling outages at each Unit.
The extended test frequency will also reduce the amount of out of service time during refueling outages for certain high head and low head safety injection valves located in the boration flow paths at each Unit by approximately 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> per Unit per refueling outage.
A review of the leakage test results for the valves identified in Tables 1, 2 and 3 shows that these valves have maintained a history of excellent performance with a few exceptions noted below:
- Fill Header Check Valve [1CH-170] failed its leak test in 2000 during the thirteenth refueling outage and was repaired (seats were lapped). Leakage rates since then have been acceptable.
- Reactor Coolant Pump 1A Thermal Barrier Cooler Inlet Check Valve [1CCR-289]
had a history of leaking more than its sister valves [1CCR-290 and 291] even though its seats were passing a 360-degree blue check. The valve was replaced during the seventeenth refueling outage in 2006, and the subsequent leak rates have been acceptable.
- Reactor Coolant System Loop Fill Header Flow Control Valve [2CHS*FCV160]
failed its leak test in 2012 during the sixteenth refueling outage and was adjusted (stem valve travel adjusted to make tighter). Subsequent leak rates have been acceptable.
This excellent valve performance supports extending the leakage test frequency for valves in Tables 1, 2 and 3 to four refueling outages based on achieving acceptable results for a least two consecutive outages prior to the submittal of this relief request.
Therefore, the proposed alternative to perform leakage testing of the valves identified in Tables 1, 2 and 3 at intervals based on valve performance in lieu of the current ASME Code paragraph ISTC-3630(a) requirement to conduct tests at least once every two years, provides reasonable assurance that the valves are operationally ready.
6. Duration of Proposed Alternative
The proposed alternative is requested for use during the remainder of the fifth 10-year inservice test interval at BVPS-1 and the fourth 10-year inservice test interval at BVPS-2. Both intervals began September 20, 2017 and end September 19, 2027.