05000354/LER-2015-004, Regarding As-Found Values for Safety Relief Valve Lift Set Points Exceed Technical Specification Allowable Limit
| ML15211A290 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 07/30/2015 |
| From: | Carr E Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N15-0161 LER 15-004-00 | |
| Download: ML15211A290 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3542015004R00 - NRC Website | |
text
LR-N15-0161 JUL 3 0 2015 PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Hope Creek Generating Station Unit 1 t>Sl(i Nuclear l.f J,'
10CFR50.73 Renewed Facility Operating License No. NPF-57 Docket No. 50-354
Subject:
Licensee Event Report 2015-004-00 In accordance with the requirements of 10 CFR 50. 73(a)(2)(i)(B),
PSEG Nuclear LLC is submitting the enclosed Licensee Event Report (LER) Number 2015-004-00, "As-Found Values for Safety Relief Valve Lift Set Points Exceed Technical Specification Allowable 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,
----
Eric S. Carr Plant Manager Hope Creek Generating Station ttm Attachment: Licensee Event Report 2015-004-00
/
LR-N15-0161 Page 2 of 2 Document Control Desk cc:
Mr. Daniel Dorman, Regional Administrator-Region I, NRC Ms. Carleen Parker, Project Manager - US NRC Justin Hawkins, NRC Senior Resident Inspector-Hope Creek (X24)
Mr. Patrick Mulligan, Manager IV Bureau of Nuclear Engineering New Jersey Department of Environmental Protection PO Box 420 Trenton, NJ 08625 10CFR50.73 Mr. Thomas MacEwen, Hope Creek Commitment Tracking Coordinator (H02)
Mr. Lee Marabella - Corporate Commitment Tracking Coordinator (N21)
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0113112017 (02-2014)
-"*flMi.t,
, tt1e NRC may not conduct or sponsor, and a person is not required to respond to, the Information collection.
- 6. LER NUMBER
- 3. PAGE YEAR I SEQUENTIAL I REVISION NUMBER NUMBER 3 OF4 2015
- - 004
- - 000 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 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.
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 14 SRV pilot stage assemblies were removed and replaced with tested spares during the refueling outage (H1R19), the extent of condition scope was satisfied.
CAUSE OF EVENT
The cause of the setpoint drift for the ten 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
All 14 SRVs were operable during Cycle 19 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.
An evaluation is In progress to assess the aggregate safety significance of the as-found condition of ten SRVs. The evaluation will review the reactor pressure vessel over-pressure protection design function of the valves; the impact of higher relief set points on other safety systems (i.e., HPCI, RCIC, and SLC); and fuel considerations. The conclusions of that evaluation will be reported in a supplement to this LER.
SAFETY SYSTEM FUNCTIONAL FAILURE An evaluation to assess the aggregate safety significance of the failure of ten SRVs is in progress. The conclusion on whether this constitutes a Safety System Functional Failure (SSFF) as defined in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Performance Indicator Guideline," will be reported in a supplement to this LER.
PREVIOUS EVENTS A review of events for the past four years at Hope Creek was performed to determine if similar events had occurred.
Similar events occurred during the 2012 (H1R17) and 2013 (H1R18) Hope Creek refueling outages when multiple SRVs were found out of the TS required limits of+/- 3%. These events were reported as LER 354/2012-004-01 (six inoperable SRVs) and LER 354/2013-007-00 (five inoperable SRVs ).
NRC FORM 366 A (02-2014)
U. S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET Hope Creek Generating Station 05000354 NARRATIVE
CORRECTIVE ACTIONS
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017
, the NRC may not conduct or sponsor, and a person is not required to respond to, the Information collection.
- 6. LER NUMBER I
SEQUENTIAL NUMBER YEAR 2015
- - 004 I REVISION NUMBER
- - 000
- 3. PAGE 4 OF4
- 1.
All 14 SRV pilot stage assemblies were removed and replaced with pre-tested, certified spare pilot valves (H1R19).
- 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.