05000247/LER-2015-002, Regarding Safety System Functional Failure Due to Fuses for Residual Heat Removal Heat Exchanger Outlet Valves That Would Not Remain Operable Under Degraded Voltage Conditions
| ML15300A059 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 10/19/2015 |
| From: | Robert Walpole Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-15-126 LER 15-002-00 | |
| Download: ML15300A059 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(1), Submit an LER, Invalid Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 2472015002R00 - NRC Website | |
text
--En tergy Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, N.Y. 10511-0249 Tel (914) 254-6710 Robert Walpole Regulatory Assurance Manager NL-15-126 October 19, 2015 U.S. Nuclear Regulatory Commission Document Control Desk 11545 Rockville Pike, TWFN-2 F1 Rockville, MD 20852-2738
SUBJECT:
Licensee Event Report # 2015-002-00, "Safety System Functional Failure Due to Fuses for Residual Heat Removal Heat Exchanger Outlet Valves That Would Not Remain Operable Under Degraded Voltage Conditions" Indian Point Unit No. 2 Docket No. 50-247 DPR-26
Dear Sir or Madam:
Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2015-002-00. The attached LER identifies an event where there was a Safety System Functional Failure due to fuses for Residual Heat Removal Heat Exchanger Outlet Valves that would not remain operable under a degraded voltage conditions resulting in both trains of RHR becoming inoperable. This condition is reportable under 10 CFR 50.73(a)(2)(v) as a safety system functional failure. This condition is also reportable under 10CFR.50(a)(2)(ii)(B) for a condition that resulted in the plant being in an unanalyzed condition that significantly degraded plant safety and under 10CFR.50(a)(2)(vii) for an event where a single cause or condition caused at least two independent trains or channels to become inoperable in a single system. This condition was recorded in the Entergy Corrective Action Program as Condition Report CR-IP2-2015-03688.
NL-15-126 Page 2 of 2 There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Regulatory Assurance at (914) 254-6710.
RW/cbr Attachment: LER-201 5-002 cc:
Mr. Daniel H. Dorman, Regional Administrator, NRC Region I NRC Resident Inspector's Office Ms. Bridget Frymire, New York State Public Service Commission
Abstract
On August 18,
- 2015, the operations shift manager entered Technical Specification (TS) 3.0.3 upon determination that the Residual Heat Removal (RER) heat exchanger outlet valves (MOV-746 and MOV-747) would not remain operable during a degraded voltage (DV) condition.
RHR heat exchanger outlet valves MOV-746 and MOV-747 are normally closed therefore their failure would result in both trains of RHR becoming inoperable.'
The RHR outlet valves are required to open during a Design Basis Accident for the RHR system to perform its safety function.
As a result of NRC inspector questioning, an evaluation of the electrical coordination calculations associated with fuses for MOV-746 and MOV-747 determined the fuses would not support continued operability during a DV condition.
The fuses were replaced with fuses that would remain operable under DV conditions and the RHR trains restored to operable status.
Direct cause was the electrical coordination calculations for MOV-746 and MOV-747 did not support operability during a DV condition.
The apparent cause was that the Industry Operating Experience (OE)
(prior NRC violations and findings) was not properly acted upon during the Focused Self-Assessment on CDBI Preparations due to an incorrect assumption.
Corrective actions included replacement of fuses, and communication to engineering personnel of the lessons learned from the event.
Initiated CR-IP2-2015-03725 recording NRC 2015 CBDI Item 103 and CR-IP2-2015-03702 recording NRC 2015 CBDI Item 104 to addresses. EOC findings and calculation updates which will be processed via Engineering Changes.
The event had no significant effect on public health and safety.
I=
(Ifmore space* isrquired, use additional copisof (Ifmore space isrequired, use additionalcopis of (Ifmore space* isrquired, use additional copisofNRCOForm 366A) (17)
Safety Significance
This event had no significant effect on the health and safety of the public. There were no actual safety consequences for the event because there were no accidents or transients during the time of the event.
A risk assessment was performed by the NRC as discussed in Inspection Report 05000247/2015-007 dated October 5, 2015.
The NRC reviewed the Entergy operability assessment and determined it was adequate.
- However, for the degraded voltage issue, the originally installed fast-acting Shawmut Type A4J30 fuses would likely actuate prior to the RHR MOVs successfully stroking open.
That condition results in the potential inoperability of both trains of low pressure injection/recirculation for longer than the TS LCO allowed outage time.
A detailed risk evaluation was performed for. that condition.
The calculated cumulative conditional core damage probability (CCDP) was determined to be 5.5E-6 with an exposure time of one year.
To account for the consequential degraded grid voltage condition, the CCDP value was multiplied by 2E-2 to approximate the probability of a LOOP given a LOCA has occurred.
The evaluation determined that the estimated increase in core damage frequency (CDF) associated with this performance deficiency is lE-7/year or very low safety significance.
The risk significance approximation is overly conservative and is considered a worst case bounding evaluation.
The risk assessment included a review for potential LERF and external events contributions.
Based on Unit 2 being a pressurized water reactor with a large dry containment, the finding screens out for LERF consideration.
Because the conditional event sequences of interest involve loss of coolant accidents, external events coincident with or contributing to these accidents would be of extremely low probability and considered beyond the plants design basis.
Accordingly, there is no external event contribution to core damage risk for this issue.