LR-N18-0065, Safety Relief Valve (SRV) As-Found Setpoint Failure

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Safety Relief Valve (SRV) As-Found Setpoint Failure
ML18169A199
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 06/18/2018
From: Casulli E
Public Service Enterprise Group
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N18-0065
Download: ML18169A199 (6)
v • d • e


Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 JUN 18 2018 LR-N 18-0065 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 NRC Docket No. 50-354 0PSEG NuclearLLC 10 CFR 50.73

Subject:

Licensee Event Report 2018-002-00, Safety Relief Valve (SRV) As-Found Setpoint Failure In accordance with 10 CFR 50. 73(a)(2)(i)(B), PSEG Nuclear LLC (PSEG) is submitting Licensee Event Report (LER) Number 2018-002-00, Safety Relief Valve (SRV) As-Found Setpoint Failure.

There are no regulatory commitments contained in this submittal. If you have any questions or require additional information, please contact Ms. Tanya Timberman at 856-339-1426.

Sincerely, J:tAtzlL Edward T. Casulli Plant Manager Hope Creek Generating Station - Licensee Event Report 2018-002-00

Page 2 LR -N 18-0065 cc:

Administrator, Region I, NRC Project Manager, NRC NRC Senior Resident Inspector, Hope Creek Mr. P. Mulligan, Chief, NJBNE Corporate Commitment Tracking Coordinator Hope Creek Commitment Tracking Coordinator 10 CFR 50.73

LR-N18-0065 Licensee Event Report 2018-002-00 Safety Relief Valve (SRV) As-Found Setpoint Failure

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (02-2018)

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LICENSEE EVENT REPORT (LER) i  )

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(See NUREG-1022, R.3 for instruction and guidance for completing this form

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1. Facility Name Hope Creek Generating Station
4. Title Safetv Relief Valve (SRV) As-Found Setooint Failure
5. Event Date
6. LER Number
7. Report Date Sequential Rev Month Day Year Year Month Day Number No.

Year 04 20 2018 2018

-002

- 00 06 18 2018 APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.

Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMS control number, the NRC may not conduct or sponsor, and a person is not required to respond to the information collection.

2. Docket Number
3. Page 05000354 1 OF3
8. Other Facilities Involved Facility Name Docket Number 05000 Facility Name Docket Number 05000
9. Operating Mode
11. This Report is Submitted Pursuant to the Requirements of 10 CFR §: (Check all that apply) 5 - Refuel D 20.2201 (b)

D 20.2201 (d)

D 20.2203(a)(1)

D 20.2203(a)(2)(i)

10. Power Level D 20.2203(a)(2)(ii)

D 20.2203(a)(2)(iii) 0%

D 20.2203(a)(2)(iv)

D 20.2203(a)(2)(v)

D 20.2203(a)(2)(vi)

Licensee Contact

!Tanya Timberman, Senior Nuclear Engineer D 20.2203(a)(3)(i)

D 50.73(a)(2)(ii)(A)

D 50.73(a)(2)(viii)(A)

D 20.2203(a)(3)(ii)

D 50.73(a)(2)(ii)(B)

D 50. 73(a)(2)(viii)(B)

D 20.2203(a)(4)

D 50. 73(a)(2)(iii)

D 50. 73(a)(2)(ix)(A)

D 50.36(c)(1 )(i)(A)

D 50.73(a)(2)(iv)(A)

D 50.73(a)(2)(x)

D 50.36(c)(1 )(ii)(A)

D 50. 73(a)(2)(v)(A)

D 73.71(a)(4)

D 50.36(c)(2)

D 50. 73(a)(2)(v)(B)

D 73.71 (a)(5)

D 50.46(a)(3)(ii)

D 50.73(a)(2)(v)(C)

D 73.77(a)(1)

D 50. 73(a)(2)(i)(A)

D 50. 73(a)(2)(v)(D)

D 73.77(a)(2)(ii) 181 50. 73(a)(2)(i)(B)

D 50.73(a)(2)(vii)

D 73.77(a)(2)(iii)

D 50.73(a)(2)(i)(C)

D Other (Specify in Abstract below or in NRC Form 366A

12. Licensee Contact for this LER rrephone Number (Include Area Code) 1856-339-1426
13. Complete One Une for each Component Failure Described in this Report Cause System Component Manufacturer Reportable To ICES Cause System Component Manufacturer Reportable To ICES 8

SB RV T020 y

14. Supplemental Report Expected Month Day Year 0 Yes (If yes, complete 15. Expected Submission Date) Z No
15. Expected Submission Date

[Abstract (Limit to 1400 spaces, i.e., approximately 14 single-spaced typewritten lines)

On April 20, 2018, Hope Creek Generating Station (HCGS) received results that the 'as-found' set-point tests 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-first refueling outage (H1R21), all fourteen SRV pilot stage assemblies were removed for testing at an offsite facility. Between April 20 and May 11, 2018, HCGS received the test results for all fourteen of the SRV pilot valve assemblies. A total of eight of the fourteen SRV pilot stage assemblies experienced set-point drift 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.

This is a condition reportable under 10 CFR 50.73(a)(2)(i)(B) as an Operation or Condition Prohibited by Technical Specifications.

The cause of the set-point drift for the eight SRV pilot stage assemblies is attributed to corrosion bonding between the pilot disc and seating surfaces, which is consistent with industry experience. This conclusion is based on previous cause evaluations and the repetitive nature of this condition at HCGS and within the BWR industry.

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NRC FORM 366 {02*2018)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 3/31/2020 (04-2017)

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET (See NUREG-1 022, R.3 for instruction and guidance for completing this form http://www. nrc.gov/readinq-rm/doc-collections/nureqs/staff/sr1 022/r3D Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Information Services Branch {T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-1 0202, {3150-01 04), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

1. FACILITY NAME
2. DOCKET NUMBER
3. LER NUMBER Hope Creek Generating Station 05000-354 NARRATIVE PLANT AND SYSTEM IDENTIFICATION General Electric-Boiling Water Reactor (BWR/4)

Main Steam-EllS Identifier {SB}*

Safety Relief Valves - EllS Identifier {SB/RV}*

YEAR 2018 SEQUENTIAL NUMBER

-002

  • Energy Industry Identification System {EllS} codes and component function identifier codes appear as {SS/CCC}

IDENTIFICATION OF OCCURRENCE Event Date: April 20, 2018 Discovery Date: April 20, 2018 CONDITIONS PRIOR TO OCCURRENCE REV NO.

-00 When the reports of the 'as-found' results were received, Hope Creek was in Operational Condition (OPCON) 5, Refuel, at 0 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-first refueling outage (H 1 R21) at Hope Creek Generating Station (HCGS), all fourteen Main Steam safety relief valves (SRV) pilot stage assemblies {SB/RV} were removed and tested at NWS Technologies. The SRVs are Target Rock Model 7567F two-stage SRVs. During the period from April 20 through May 11, 2018, 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 eight 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 eight SRVs not meeting the TS requirements are as follows:

Valve ID As Found TS Lift Setting Acceptable Band

% Difference (psig)

(psig)

(psig)

Actual F013B 1210 1130 1096.1-1163.9 7.10%

F013D 1191 1130 1096.1 -1163.9 5.40%

F013F 1146 1108 1074.8-1141.2 3.40%

F013G 1197 1120 1 086.4 -1153.6 6.90%

F013H 1200 1108 1074.8-1141.2 8.30%

F013L 1155 1120 1086.4-1153.6 3.10%

F013M 1161 1108 1074.8-1141.2 4.80%

F013P 1199 1120 1086.4-1153.6 7.10%

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 +I-3%. Action (a) of TS 3.4.2. 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 />." Therefore, this is a condition reportable under 10 CFR 50.73(a)(2)(i)(B) as an Operation or Condition Prohibited by TS.

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NRC FORM 3668 (02-2018)

Page 2 of 3

NRC FORM 366A (04-2017)

U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 3/31/2020

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LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET (See NUREG-1 022, R.3 for instruction and guidance for completing this form http://www. nrc.qov/readinq*rm/doc-collections/nureqs/staff/sr1 022/r3D Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

1. FACILITY NAME
2. DOCKET NUMBER
3. LER NUMBER YEAR Hope Creek Generating Station 05000-354 SEQUENTIAL NUMBER REV NO.

2018

-002

-00 NARRATIVE DESCRIPTION OF OCCURRENCE (Continued)

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. However, since all fourteen SRV pilot stage assemblies were removed and replaced with tested spares during the refueling outage (H1 R21 ), the extent of condition scope was satisfied.

CAUSE OF EVENT The cause of the set-point drift for the eight SRV pilot stage assemblies is attributed to corrosion bonding between the pilot disc and seating surfaces, which is consistent with industry experience. This conclusion is based on previous cause evaluations and the repetitive nature of this condition at HCGS and within the BWR industry.

SAFETY CONSEQUENCES AND IMPLICATIONS There were no instances during cycle 21 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 Limit, providing reasonable assurance that accident analysis conclusions would remain valid. The industry 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.

Since the eight as-found setpoint SRVs are within their Maximum Allowable Percent Increase (MAP I) above SRV nominal setpoint criteria established in GE document NEDC-32511 P, "Safety Review for Hope Creek Generating Station Safety/Relief Valve Tolerance Analysis", the SRVs are bounded by their MAPI value and no formal Technical Evaluation is required.

SAFETY SYSTEM FUNCTIONAL FAILURE A review of this condition and a previous technical evaluation documents this is not a functional failure, therefore it was determined that a Safety System Functional Failure (SSFF) as defined in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Performance Indicator Guideline," did not occur.

PREVIOUS EVENTS Similar events occurred during the 2015 (H1R19) and 2016 (H1R20) 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 (ten inoperable SRVs) and LER 354/2016-003-00 (ten inoperable SRVs).

CORRECTIVE ACTIONS

1.

All 14 SRV pilot stage assemblies were removed and replaced with pre-tested, certified spare pilot valves (H1 R21).

2.

Evaluate options for the replacement of the currently installed Target Rock two-stage SRVs with a design that eliminates setpoint drift events exceeding +/-3% and improve SRV reliability. The replacement schedule will be developed after a suitable valve is identified.

COMMITMENTS There are no regulatory commitments contained in this LER.

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NRC FORM 366A (06-2016)

Page 3 of :3

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