05000354/LER-2018-003, Feedwater Isolation Valve Leakage Exceeded Technical Specification Limit
| ML18169A307 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/18/2018 |
| From: | Casulli E Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N18-0068 LER 2018-003-00 | |
| Download: ML18169A307 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3542018003R00 - NRC Website | |
text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 JUN
.. IS 2018 LR-N 18-0068 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Hope Creek Generating Station 0PSEG Nuclear LLC 10CFR50.73 Renewed Facility Operating License No. NPF-57 Docket No. 50-354
Subject:
Licensee Event Report 2018-003-00, Feedwater Isolation Valve Leakage Exceeded Technical Specification Limit In accordance with 10 CFR 50. 73(a)(2)(i)(B), PSEG Nuclear LLC is submitting Licensee Event Report (LER) Number 2018-003-00, "Feedwater Isolation Valve Leakage Exceeded Technical Specification Limit."
If you have any questions or require additional information, please contact Mr. Thomas MacEwen at (856) 339-1097.
There are no regulatory commitments contained in this letter.
Sincerely,
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- fen" rd T. Casulli Plant Manager Hope Creek Generating Station ttm
Attachment:
Licensee Event Report 2018-003-00
LR-N18-0068 Page 2 cc:
Mr. Daniel Dorman, Regional Administrator-Region I, NRC Mr. Jim Kim, Project Manager-US NRC 10CFR50.73 Mr. Justin Hawkins, NRC Senior Resident Inspector-Hope Creek (X24)
Mr. Patrick Mulligan, Manager IV, NJBNE Mr. Thomas MacEwen, Hope Creek Commitment Tracking Coordinator (H02)
Mr. Lee Marabella, Corporate Commitment Tracking Coordinator (N21)
NRCFORM366 U.S. NUCLEAR REGULATORY COMMISSION (02-2018)
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LICENSEE EVENT REPORT (LER)
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- 1. Facility Name Hope Creek Generating Station
- 4. Title APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020
, the NRC may not conduct or sponsor, and a person is not reouired to respond to the information collection.
- 2. Docket Number
-Page 05000-354 1 OF3 Feedwater Isolation Valve Leakaae Exceeded Technical Specification Limit
- 5. Event Date
- 6. LER Number
- 7. Report Date
- 8. Other Facilities Involved Sequential Rev Facility Name Docket Number Month Day Year Year Month Day Year Number No.
05000 04 2018 2018
- - 003
- - 00 06 18 2018 Facility Name Docket Number 18 05000
- 9. Operating Mode 08 31 2018 Abstract (Limit to 1400 spaces, i.e., approximately 14 single-spaced typewritten lines)
On April 18, 2018, with Hope Creek Generating Station (HCGS) in a planned refueling outage, HCGS performed a required surveillance test of the long term seal of the feedwater lines. The test criteria could not be met due to leakage past feedwater isolation valve H1 AE -AE-HVF032B. The valve is sealed with a water seal from the High Pressure Reactor Coolant (HPCI) system, or Reactor Core Isolation Cooling (RCIC) system to form a long-term seal boundary of the feedwater lines. The valve is tested per Technical Specification 4.6.1.2.d to verify a maximum leak rate of 10 gpm at a test pressure of 55.7 psig. During the test, a test pressure of 44 psig was the highest pressure that could be obtained, which does not meet the acceptance criterion.
These conditions are being reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by plant Technical Specifications.
NRC FORM 366 {02*2018)
PLANT AND SYSTEM IDENTIFICATION
General Electric-Boiling Water Reactor (BWR/4)*
Feedwater System (SJ) - EllS Identifier {SJ/ISV}
High Pressure Coolant Injection (BJ ) - EllS Identifier {BJ}
Reactor Core Isolation Cooling (BN) - EllS Identifier {BN}
Secondary Containment (NG) - EllS Identifier {NG}
YEAR 2018 SEQUENTIAL NUMBER
- - 003
- Energy Industry Identification System {EllS} codes and component function identifier codes appear as {SS/CCC}
IDENTIFICATION OF OCCURRENCE Event Dates: April 18, 2018 Discovery Dates: April 18, 2018 CONDITIONS PRIOR TO OCCURRENCE Hope Creek was shut down for Refueling Outage H1 R21 in Operational Condition (OPCON) 5-Refueling Operations.
DESCRIPTION OF OCCURRENCE REV NO.
- - 00 On April 18, 2018, with Hope Creek Generating Station (HCGS) in a planned refueling outage, HCGS performed a required surveillance test of the long term seal of the feedwater lines. The test criteria could not be met due to leakage past feedwater system {SJ} isolation valve H1 AE -AE-HVF032B. The valve is sealed with a water seal from the High Pressure Coolant Injection (HPCI) {BJ} system, or Reactor Core Isolation Cooling (RCIC) {BN} system, to form a long term seal boundary of the feedwater lines. The valve is tested with water at a test pressure of 55.7 psig to ensure the seal boundary will prevent bypass leakage. During the test, a test pressure of 44 psig was the highest pressure that could be obtained, which does not meet the acceptance criterion.
The H1 AE -AE-HV-F032B is 24 inch anchor darling "Y" type swing check valve with a Limitorque SMB-4 motor operator to assist in maintaining the valve closed.
Technical Specification 4.6.1.2.d limits the total combined leakage rate to 10 gpm, or less, for all the containment isolation valves which form the boundary for the long term seal of the feedwater lines, when tested at 55.7 psig. Based on the cause of the failure, and the maintenance history of the valve, it was concluded that the condition is reportable as a condition prohibited by Technical Specifications under 10 CFR 50.73(a)(2)(i)(B).
CAUSE OF EVENT
An internal inspection was performed on the H1 AE -AE-HV-F032B under work order 60138581 in H1 R21. The following was identified during this inspection:
- - One bore on the hinge arm was larger than the acceptance criteria by an average of 0.004 inches. (2.039 inches vs 2.029 - 2.035 inches)
- - The hinge pin was worn and undersized by 0.002 inches. (1.997 inches vs 2.000 +1- 0.001 inches)
- - The gap between the hinge arm and disc was smaller by 0.0325 inches (0.030 inches vs 0.0625 inches)
The combination of the wear on the hinge arm bores, the wear on the hinge pin, and the smaller gap between the hinge arm and disc prevented the valve disc from properly seating, which resulted in the high leak rate.
NRC FORM 3668 (02-2018)
FJage 2 of 3 (04-2017)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 3/31/2020
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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/r3l)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. LER NUMBER Hope Creek Generating Station 05000-354
SAFETY CONSEQUENCES AND IMPLICATIONS
YEAR 2018 SEQUENTIAL NUMBER
- - 003 The long term feedwater seal is established following a Loss of Coolant Accident (LOCA) by manually aligning the RCIC and/or HPCI jockey pumps to the feedwater lines between the inboard and outboard containment isolation REV NO.
- - 00 valves. H1AE -AE-HV-F0328 is the outboard containment isolation valve on one of two feedwater supply lines. The purpose of the feedwater seal, as described in the UFSAR, is to establish a water seal of the feedwater penetrations to eliminate bypass leakage. The seal is established following a LOCA, and is to be maintained for a minimum of 30 days.
The purpose of the leak rate testing is to verify that the leakage is within the capability of the system to maintain the seal for the 30 day minimum.
An evaluation is in progress to determine if the as-found leakage would have affected the ability to maintain the feedwater seal for the 30 day minimum. A supplement to this LER will be submitted when the results of the evaluation are available.
PREVIOUS EVENTS A review of Licensee Event Reports and the corrective action program for the past three years identified no LERs issued for similar conditions.
CORRECTIVE ACTIONS
The F0328 check valve was opened and inspected. Corrective maintenance was performed to address the dimensional clearances and deficiencies identified in the Cause of Event section above. A satisfactory leak rate test was performed following the corrective maintenance.
COMMITMENTS
This LER contains no regulatory commitments.
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