05000354/LER-2018-003-01, Feedwater Isolation Valve Leakage Exceeded Technical Specification Limit

Feedwater Isolation Valve Leakage Exceeded Technical Specification Limit

ML18255A232
Person / Time
Site:	Hope Creek
Issue date:	09/12/2018
From:	Casulli E Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N18-0093 LER 2018-003-01
Download: ML18255A232 (5)

LER-2018-003, Feedwater Isolation Valve Leakage Exceeded Technical Specification Limit

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3542018003R01 - NRC Website

text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0PSEG Nuclear LLC LR-N 18-0093 SEP 12 2018 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC. 20555-001 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 Docket No. 50-354 10CFR50.73

Subject:

Supplemental Licensee Event Report 2018-003-01, Feedwater Isolation Valve Leakage Exceeded Technical Specification Limit

Reference:

PSEG Letter LR-N18-0068, dated June 18, 2018 Licensee Event Report 2018-003-00 In accordance with 10 CFR 50.73(a)(2)(i)(B), PSEG Nuclear LLC is submitting Supplemental Licensee Event Report (LER) Number 2018-003-01, "Feedwater Isolation Valve Leakage Exceeded Technical Specification Limit." The Reference LER stated that Hope Creek Generating Station would supply a supplement to the LER with the results of an evaluation to determine if the leakage would have prevented the fulfillment of a safety function. The results of the evaluation are communicated in the LER supplement attached to this letter.

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,

~TUI Edward T. Casulli Plant Manager Hope Creek Generating Station Attachment: Supplemental Licensee Event Report 2018-003-01

LR-N18-0093 Page 2 cc:

Mr. David Lew, Regional Administrator-Region I, NRC Mr. James 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 APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0313112020 (04-2018)

, the NRC may not conduct or sponsor, and a oerson is not reauired to resoond to the information collection.

3.Page Hope Creek Generating Station 05000-354 1 OF3

4. Title Feedwater Isolation Valve Leakaqe 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 18 2018 2018

- 003

- 01 09 12 2018 Facility Name Docket Number 05000

9. Operating Mode

~bstract (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 H1AE -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 Surveillance Requirement 4.6.1.2.g 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.

This condition is being reported in accordance with 10 CFR 50. 73(a)(2)(i)(B) as a condition prohibited by plant Technical Specifications.

NRC FORM 366 {04-2018)

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: April18, 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.

- 01 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 H1AE -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. At the maximum water test rig pressure, the test rig's downstream pressure was only able to achieve approximately 44 psig, 11.7 psig less than the required 55.7 psig to perform the test.

Although no flow rate was achieved or directly measured, the test was considered failed due to not being able to meet procedural test acceptance criterion.

The H1AE -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 3.6.1.2.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 +/- 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 (04-2018)

Page 2 of 3 (04-2018)

U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET (See NUREG-1022, R.3 for instruction and guidance for completing this form htto://www.nrc.gov/readinq-rm/doc-collections/nureos/staff/sr1022/r3/)

APPROVED BY OMB: NO. 3150-0104 EXPIRES: 3/31/2020

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

1. FACILITYNAME

2. DOCKET NUMBER

3. L.ER 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.

- 01 valves. H1AE -AE-HV-F032B 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.

A Technical Evaluation was performed to determine the impact on the ability to maintain a leak seal for the 30 day minimum. The evaluation of test data indicates leakage through the F032B would not have posed a challenge to the ability to establish and maintain the required feedwater seal for 30 days post-LOCA.

SAFETY SYSTEM FUNCTIONAL FAILURE A review of this condition, and the associated technical evaluation, 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. This event did not prevent the ability of a system to fulfill its safety function to either shutdown the reactor, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident.

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 F032B 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. Page 3 of 3