05000354/LER-2013-002, Regarding Reactor Scram Due to Degrading Condenser Vacuum

Regarding Reactor Scram Due to Degrading Condenser Vacuum

ML13224A304
Person / Time
Site:	Hope Creek
Issue date:	08/08/2013
From:	Carr E Public Service Enterprise Group
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
LR-N13-0162 LER 13-002-00
Download: ML13224A304 (6)

LER-2013-002, Regarding Reactor Scram Due to Degrading Condenser Vacuum

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2) 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)(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)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3542013002R00 - NRC Website

text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 AUG 0 B 2013 LR-N13-0162 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001

Subject:

Hope Creek Generating Station Unit 1 Renewed Facility Operating License No. NPF-57 Docket No. 50-354 Licensee Event Report 2013-002-00 PSEG Nuclear LLC 10CFR50.73 In accordance with 10 CFR 50.73(a)(2)(iv)(A), PSEG Nuclear LLC is submitting Licensee Event Report (LER) Number 2013-002-00, "Reactor Scram due to Degrading Condenser Vacuum."

Should you have any questions concerning this letter, please contact Mr. Paul Bonnett at (856) 339-1923.

No regulatory commitments are contained in the LER.

Sincerely, Eric S. Carr Plant Manager Hope Creek Generating Station Attachment: Licensee Event Report 2013-002-00

Document Control Desk LR~N13-0162 Page 2 cc:

Mr. W. Dean, Regional Administrator - Region I U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 Mr. J. Hughey, Project Manager U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 B1A 11555 Rockville Pike Rockville, MD 20852 USNRC Senior Resident Inspector - Hope Creek (X24)

P. Mulligan, Manager Bureau of Nuclear Engineering New Jersey Department of Environmental Protection PO Box 420 Me 33-01 33 Arctic Parkway Trenton, NJ 08625 Hope Creek Commitment Tracking Coordinator (H02)

Corporate Commitment Tracking Coordinator (N21)

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010)

, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the Information collection.

3. PAGE Hope Creek Generating Station 05000354 1 OF4

4. TITLE Reactor Scram due to Degrading Condenser Vacuum

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED SEQUENTIAL REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.

MONTH DAY YEAR N/A FACILITY NAME DOCKET NUMBER 06 12 2013 2013 - 002 - 00 08 08 2013 N/A

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) o 20.2201(b) o 20.2203(a)(3)(i) o 50.73(a)(2)(i)(C) o 50.73(a)(2)(vii) 1 o 20.2201(d) o 20.2203(a)(3)(ii) o 50.73(a)(2)(ii)(A) o 50.73(a)(2)(vili)(A) o 20.2203(a)(1)

D 20.2203(a)(4)

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

D 50.73(a)(2)(viii)(B)

D 20.2203(a)(2)(i)

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

D 50.73(a)(2)(iii)

D 50.73(a)(2)(ix)(A)

10. POWER LEVEL D 20.2203(a)(2)(ii)

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

[gI50.73(a)(2)(iv)(A)

D 50.73(a)(2)(x)

D 20.2203(a)(2)(iii)

D 50.36(c)(2)

D 50.73(a)(2)(v)(A)

D 73.71(a)(4)

D 20.2203(a)(2)(iv)

D 50.46(a)(3)(ii)

D 50.73(a)(2)(v)(B)

D 73.71(a)(5) 100 D 20.2203(a)(2)(v)

D 50.73(a)(2)(i)(A)

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

D OTHER D 20.2203(a)(2)(vi)

D 50.73(a)(2)(i)(B) o 50.73(a)(2)(v)(D)

Specify in Abstract below or in This event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) as a valid manual actuation of RPS and manual initiation of the RCIC system.

Additional Background

3. PAGE 30F4 On May 7, 2013, the 'B' CW discharge valve (DA-HV-2152B) failed to stroke from OPEN-FULL to OPEN-MID while attempting to remove the 'B' CW pump from service for maintenance. The CW discharge valves are model Triton XR-70, 84 inch, butterfly valves manufactured by PRATT. The valves are operated hydraulically by a hydraulic power unit. Initial troubleshooting revealed the valve was unable to be stroked closed. An Operational and Technical Decision Making (OTDM) document determined the acceptability of continued operation with the degraded discharge valve until it could be repaired in the fall refueling outage.

The evaluation acknowledged the risk of a reactor scram on low condenser vacuum, especially during the summer months, however it concluded there would be no adverse impact to plant components or nuclear safety.

CAUSE OF EVENT

A root cause evaluation is in progress. The results of the evaluation will be published in a supplement to this LER.

SAFETY CONSEQUENCES AND IMPLICATIONS

There were no nuclear safety consequences associated with this event. All control rods fully inserted following the initiation of the manual reactor scram. There were no automatic initiations of safety systems, and immediate actions performed by the operators were adequate and appropriate in placing and maintaining the reactor in a safe shutdown condition. The loss of condenser vacuum classified the event as an unplanned scram with complications in accordance with NEI 99-02; however, it is concluded that the safety significance of this event was low and the event did not pose a threat to the health and safety of the public or plant personnel.

SAFETY SYSTEM FUNCTIONAL FAILURE A review of this event determined that a Safety System Functional Failure (SSFF) as defined in NEI 99-02, "Regulatory Assessment Performance Indicator Guidelines," 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 root cause evaluation is in progress. Previous events will be determined after the root cause evaluation is complete.

NRC FORM 366 (10-2010)

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CORRECTIVE ACTIONS

1.

The 'B' circulating water discharge valve (OA-HV-21S2B) was repaired during the forced outage.

2.

The components most likely to have caused the 'B' circulating water pump trip were replaced.

3.

A root cause evaluation is in progress. Corrective actions will be published in a supplement to this LER.

COMMITMENTS

This LER contains no regulatory commitments.

NRC FORM 366 (10-2010)

PRINTED ON RECYCLED PAPER