05000354/LER-2019-002, Safety Relief Valve (SRV) As-found Set-point Failures
| ML20006E541 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 01/06/2020 |
| From: | Poorman S Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N19-0110 LER 2019-002-00 | |
| Download: ML20006E541 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 3542019002R00 - NRC Website | |
text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 PSEG Nuclear LLC LR-N19-0110 JAN O 6 2020 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001
Subject:
Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 Docket No. 50-354 Licensee Event Report 2019-002-00, Safety Relief Valve (SRV) As-found Set-point Failures 10CFR50.73 In accordance with 10 CFR 50.73(a)(2)(i)(B), PSEG Nuclear LLC is submitting Licensee Event Report (LER) Number 2019-002-00, "Safety Relief Valve (SRV) As-found Set-point Failures."
If you have any questions or require additional information, please contact Mr. Francis D. Possessky at (856) 339-1160.
There are no regulatory commitments contained in this letter.
s~~
Steven R. Poorman Plant Manager Hope Creek Generating Station Attachment: Licensee Event Report 2019-002-00 cc:
Regional Administrator-Region I, NRG US NRG NRR Project Manager - Hope Creek US NRG Senior Resident Inspector - Hope Creek NJ Department of Environmental Protection, Bureau of Nuclear Engineering Commitment Coordinator, Hope Creek Generating Station Corporate Commitment Coordinator, PSEG Nuclear LLC
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BYOMB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2018)
, the NRC may not conduct or sponsor, and a person is not required to httQ://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1 022/r30 respond to, the information collection.
- 3. PAGE Hope Creek Generating Station 05000354 1 OF 3
- 4. TITLE Safety Relief Valve (SRV) As-found Set-point Failures
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED I
SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
MONTH DAY YEAR 05000 11 07 2019 2019 - 002
- - 00 01 06 2020 FACILITY NAME DOCKET NUMBER 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 4 - Cold Shutdown r 20.2201(b) r 20.2203(a)(3)(i)
- 50. 73(a)(2)(ii)(A)
- 50. 73(a)(2)(viii)(A) r 20.2201(d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) r 20.2203(a)(1) r 20.2203(a)(4) r 50.73(a)(2)(iii) r 50.73(a)(2)(ix)(A) r 20.2203(a)(2)(i) r 50.36(c)(1)(i)(A) r 50.73(a)(2)(iv)(A) r 50.73(a)(2)(x)
- 10. POWER LEVEL r 20.2203(a)(2)(ii) r 50.36(c)(1)(ii)(A) r 50.73(a)(2)(v)(A) r 73.71(a)(4) 000 r 20.2203(a)(2)(iii) r 50.36(c)(2) r 50.73(a)(2)(v)(B) r 73.71(a)(5) 20.2203(a)(2)(iv) r 50.46(a)(3)(ii) r 50.73(a)(2)(v)(C) 73.77(a)(1) r 20.2203(a)(2)(v) r 50.73(a)(2)(i)(A) r 50.73(a)(2)(v)(D) r 73.77(a)(2)(i) r 20.2203(a)(2)(vi)
P' 50.73(a)(2)(i)(B) r 50.73(a)(2)(vii) r 73.77(a)(2)(ii) r 50.73(a)(2)(i)(C)
OTHER Specify in Abstract belcmor in NRG FOfm 366A
- 12. LICENSEE CONTACT FOR THIS LER LICENSEE CONTACT TELEPHONE NUMBER (Include Area Code)
Francis D. Possessky, Regulatory Compliance Specialist (856) 339-1160 CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX B
SB RV T020 y
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR r YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
P' NO SUBMISSION DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On November 08, 2019, Hope Creek Generating Station (HCGS) received results that the second 'as-found' set-point test for safety relief valve (SRV) pilot stage assemblies had exceeded the lift setting tolerance prescribed in Technical Specification (TS) 3.4.2.1. The TS requires the SRV lift settings to be within +/- 3% of the nominal set-point value.
During the twenty-second refueling outage (H1 R22), all fourteen SRV pilot stage assemblies were tested at an offsite facility. Between October 22 and November 25, 2019, HCGS received the test results for all fourteen of the SRV pilot valve assemblies. A total of six of the fourteen SRV pilot stage assemblies were outside of the TS 3.4.2.1 specified values. All of the valves failing to meet the limits were Target Rock Model 7567F two-stage SRVs.
Exceeding the set points for five of the six SRV pilot stage assemblies is attributed to corrosion bonding between the pilot discs and seating surfaces, which is consistent with industry experience. The cause of exceeding the sixth SRV set-point was pilot failure or spindle/disc contact wear.
NRC FORM 366 (04-2018)
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor (BWR/4)
Main Steam - EIIS Identifier {SB}*
Safety Relief Valves - EIIS Identifier {SB/RV}*
05000354 APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 03/31/2020
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. LER NUMBER YEAR I SEQUENTIAL I REVISION NUMBER NUMBER 2019
- - 002
- - 00
- Energy Industry Identification System {EIIS} codes and component function identifier codes appear as {SS/CCC}
IDENTIFICATION OF OCCURRENCE Event Date: November 7, 2019 Discovery Date: November 8, 2019 CONDITIONS PRIOR TO OCCURRENCE When the reports of the 'as-found' results were received, Hope Creek was in Operational Condition (OPCON) 4, Cold Shutdown, at O percent rated thermal power. No other structures, systems or components that could have contributed to the event were inoperable at the time of the event.
DESCRIPTION OF OCCURRENCE During the twenty-second refueling outage (H1 R22) at Hope Creek Generating Station (HCGS), all fourteen Main Steam safety relief valves (SRV) pilot stage assemblies {SB/RV} were tested at NWS Technologies. The SRVs are Target Rock Model 7567F two-stage SRVs. During the period from October 22 through November 25, 2019, HCGS received the results of the 'as-found' set pressure testing required by Technical Specification (TS) Surveillance Requirement (SR) 4.4.2.2. A total of six of the fourteen SRV pilot stage assemblies had set-point drift outside of the required TS 3.4.2.1 tolerance values of +/-3% of nominal value.
The 'as-found' test results for the six SRVs not meeting the TS requirements are as follows:
Valve ID As Found TS Lift Setting Acceptable Band
% Difference (psig)
(psig)
(psig)
Actual F013A 1187 1130 1096.1-1163.9 5.04%
F013C 1209 1130 1096.1-1163.9 6.99%
F013D 1166 1130 1096.1-1163.9 3.19%
F013H 1194 1108 1074.8-1141.2 7.76%
F013J 1191 1120 1086.4 - 1153.6 6.34%
F013K 1055 1108 1074.8-1141.2
- - 4.78%
Technical Specification (TS) 3.4.2.1 requires that the safety function of at least 13 of 14 SRVs be operable with a specified code safety valve function lift setting, within a tolerance of+/- 3%. Action (a) of TS 3A.2.1 specifies "With the safety valve function of two or more of the above listed fourteen safety/relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." This is a condition reportable under 10 CFR 50.73(a)(2)(i)(B).
The extent of condition for this event is to expand the scope of the SRV Group 1 valve testing, per ASME OM Code Section 1-1320 for Class 1 Pressure Relief Valves. All fourteen SRV pilot stage assemblies were removed and replaced with tested spares during the refueling outage (H1R22), therefore the extent of condition was satisfied.
CAUSE OF EVENT
APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 03/31/2020
, the NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. LER NUMBER YEAR I SEQUENTIAL I REVISION NUMBER NUMBER 2019
- - 002
- - 00 Exceeding the set points for five of the six SRV pilot stage assemblies is attributed to corrosion bonding between the pilot discs and seating surfaces, which is consistent with industry experience. This conclusion is based on previous causal evaluations, the second lift test being within the acceptable band and the repetitive nature of this condition at HCGS and within the BWR industry.
The cause of exceeding the 'K' SRV set point was pilot failure or spindle/ disc contact wear. This conclusion is based on a review and disposition of Electrical Power Research Institute (EPRI) TR-105872, "Safety and Relief Valve Testing and Maintenance Guide", Section 3.1, Failure Modes and Failure Cause Analysis.
SAFETY CONSEQUENCES AND IMPLICATIONS
There were no instances during Cycle 22 that resulted in any of the fourteen SRVs being declared inoperable and there were no events during that cycle that required operation of the SRVs. All SRVs lifted well below the Safety Limi(
providing reasonable assurance that accident analysis conclusions would remain valid. The BWR Owners Group has recognized that corrosion bonding occurs during the operating cycle. Once an SRV lifts, the corrosion bond breaks and subsequent openings occur very close to the set point as demonstrated during testing.
Five of the six as-found set-point SRVs are within their Maximum Allowable Percent Increase (MAPI) above SRV nominal set-point criteria established in GE document NEDC-32511 P, "Safety Review for Hope Creek Generating Station Safety/Relief Valve Tolerance Analysis"; these SRVs are bounded by their MAPI value.
The 'A' SRV as-found set-point was outside the MAPI above SRV nominal set-point criteria established in GE document NEDC-32511 P, "Safety Review for Hope Creek Generating Station Safety/Relief Valve Tolerance Analysis", and a Technical Evaluation was performed. This evaluation concluded that the 'A' SRV set-point drift of 5.04% did not adversely affect the design function of the 'A' Main Steam Line or any segment of the 'A' SRV discharge line.
PREVIOUS EVENTS Similar events occurred during the 2015 (H1 R 19), 2016 (H1 R20) and 2018 (H1 R21) Hope Creek refueling outages when multiple SRVs were found out of the TS required limits of+/- 3%. These events were reported as LER 354/2015-004-00, LER 354/2016-003-00 and LER 354/2018-002-01.
CORRECTIVE ACTIONS
- 1.
All 14 SRV pilot stage assemblies were removed and replaced with pre-tested, certified spare pilot valves in H1 R22. The replacement pilot valve discs were coated with a platinum coating applied using Plasma Enhanced Magnetron Sputtering (PEMS) to prevent corrosion bonding, which is the most current and effective method endorsed by the BWROG.
- 2.
Replace the remaining installed Target Rock two-stage SRVs with the three-stage SRV model evaluated and implemented through Design Change Package 80122932 during the next Hope Creek Refuel Outages.
COMMITMENTS
There are no regulatory commitments contained in this LER. Page 3 of 3