05000293/LER-1917-007, Regarding Potential Inoperability of Safety Relief Valve 3A

Regarding Potential Inoperability of Safety Relief Valve 3A

ML17181A169
Person / Time
Site:	Pilgrim
Issue date:	06/22/2017
From:	Perkins E Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2.17.047 LER 17-007-00
Download: ML17181A169 (6)

LER-1917-007, Regarding Potential Inoperability of Safety Relief Valve 3A

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)
2931917007R00 - NRC Website

text

~Entergx June 22, 2017 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station

, 600 Rocky Hill Road Plymouth, MA 02360

SUBJECT:

Licensee Event Report 2017-007-00, Potential lnoperability of Safety Relief Valve3A Pilgrim Nuclear Power Station Docket No. 50-293 Renewed License No. DPR-35 LETTER NUMBER:

2.17.047

Dear Sir or Madam:

The enclosed Licensee Event Report 2017-007-00, Potential lnoperability of Safety Relief Valve 3A, is submitted in accordance with Title 10 Code of Federal Regulations 50. 73.

If you have any questions or require additional information, please contact me at (508) 830-8323.

There are no regulatory commitments contained in this letter.

Sincerely,

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Attachment:

Licensee Event Report 2017-007-00, Potential lnoperability of Safety Relief Valve 3A (3 Pages)

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station cc:

Mr. Daniel H. Dorman Regional Administrator, Region I U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd., Suite 100 King of Prussia, PA 19406-2713 Mr. John Lamb, Senior Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8C2A Washington, DC 20555 USNRC Senior Resident Inspector Pilgrim Nuclear Power Station Letter No. 2.17.047 Page 2 of 2

/

Attachment Letter Number 2.17.047 Licensee Event Report 2017-007-00 Potential lnoperability of Safety Relief Valve 3A (3 Pages)

NRCFORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0313112020 (04-2017) htto://www.nrc.gov/reading-rm/doc-collections/nuq~s/staff/sr1022/r3D the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. PAGE Pilgrim Nuclear Power Station 05000293 1 OF3

4. TITLE Potential lnoperability of Safety Relief Valve 3A

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

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

05000 N/A 04 24 2017 2017

- 007 00 06 22 2017 FACILITY NAME DOCKET NUMBER N/A 05000 N/A

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: 'Check all that apply)

N D 20.2201(b)

D 20.2203(a)(3)(i)

D 50.73(a)(2)(ii)(A)

D 50.73(a)(2)(viii)(A)

D 20.2201(d)

D 20.2203(a)(3)(ii)

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

D 50. 73(a)(2)(viii)(B)

D 20.2203(a)(1)

D 20.2203(a)(4)

D 50.73(a)(2)(iii)

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

D 20.2203(a)(2)(i)

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

D 50.73(a)(2)(iv)(A)

D 50.73(a)(2)(x)

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

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

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

D 73.71(a)(4)

D 20.2203(a)(2)(iii)

D 50.36(c)(2) 181 50.73(a)(2)(v)(B)

D 73.71(a)(5) 0 D 20.2203(a)(2)(iv)

D 50.46(a)(3)(ii)

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

D 73.77(a)(1)

D 20.2203(a)(2)(v)

D 50.73(a)(2)(i)(A) 181 50.73(a)(2)(v)(D)

D 73.77(a)(2)(i)

D 20.2203(a)(2)(vi) 181 50.73(a)(2)(i)(B)

D 50.73(a)(2)(vii)

D 73.77(a)(2)(ii)

D 50.73(a)(2)(i)(C) 0 OTHER Specify in Abstract below or in NRC Form 366A

12. LICENSEE CONTACT FOR THIS LER LICENSEE CONTACT r1LEPHONE NUMBER (Include Area Code)

IMr. Everett P. Perkins, Jr.- Regulatory Assurance Manager 1508-830-8323 CAUSE SYSTEM COMPONENT MANU-REPORTABLE

CAUSE

SYSTEM COMPONENT MANU-REPORTABLE FACTURER TOEPIX FACTURER TOEPIX B

SB RV T020 y

14. SUPPLEMENTAL REPORT EXPECTED

15. EXPECTED MONTH DAY YEAR D YES (If yes, complete 15. EXPECTED SUBMISSION DATE)

[8J NO SUBMISSION DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)

On April 24, 2017, during Refueling Outage 21 while performing testing on the Pilgrim Nuclear Power Station (PNPS) Safety/Relief Valves, a high resistance was measured across the solenoid pilot valve coil of SV203-3A. This solenoid pilot valve was replaced during Refueling Outage 21. After the solenoid pilot valve was removed it was transported to an offsite vendor for additional testing.

PNPS will continue to follow the testing performed by our offsite vendor and provide any additional information, as appropriate, to the NRC.

PNPS currently believes that this event is reportable under 10 CFR 50.73(a)(2)(i)(B), as a condition prohibited by Technical Specifications and also, potentially reportable under 50.73(a)(2)(v)(B) and 50.73(a)(2)(v)(D), a condition that could have prevented fulfillment of a safety function needed to remove residual heat and mitigate the consequences of an accident.

This event posed no threat to public health and safety.

NRC FORM 366 (11-2015)

BACKGROUND SEQUENTIAL NUMBER

- 007 REV NO.

- 00 The 2-stage pilot operated safety relief valve consists of two principle assemblies: a pilot valve section (top works) and the main va!ve section. The pilot valve section (first stage) is the pressure sensing and control element and the main valve (second stage) provides the pressure relief function. The first stage consists of a pilot-stabilizer disc assembly. The pilot is the pressure sensing member to which the stabilizer disc movement is coupled. Though not mechanically connected, a light spring keeps the stabilizer in contact with the pilot. A pilot preload spring permits set point adjustment of the valve and provides pilot seating force. The second or main stage consists essentially of a large piston which includes the main valve disc, the main valve chamber, and a preload spring.

PNPS has four safety relief valves. Each of the four relief valves is equipped with an accumulator and check valve arrangement. These accumulators are provided to assure that the valves can be held open following failure of the nitrogen supply to the accumulators, and are sized to contain sufficient nitrogen for a minimum of 20 valve operations for each safety relief valve. Bottled gas can be used to manually recharge the accumulators associated with two safety relief valves. This capability was installed to address a potential loss of normal nitrogen supply to the accumulators which was identified during seismic reviews.

EVENT DESCRIPTION

On April 24, 2017, while performing testing on the Pilgrim Nuclear Power Station (PNPS) safety relief valves a high resistance was measured across the solenoid valve coil circuit of SV203-3A.

CAUSE OF THE EVENT

The degradation mechanism has been determined to be the solenoid pilot valve coil with high electrical resistance.

CORRECTIVE ACTIONS

Removed and replaced solenoid pilot valve assembly for SV203-3A.

Page 2 of 3 (04-2017)

U.S. NUCLEARREGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 3/31/2020

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I 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/nureas/staff/sr1022/r3/)

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

3. LER NUMBER YEAR Pilgrim Nuclear Power Station 05000-293 2017

SAFETY CONSEQUENCES

SEQUENTIAL NUMBER

- 007 REV NO.

- 00 There are no consequences to the general safety of the public, nuclear safety, industrial safety and radiological safety from this event. There was a potential inoperability of the Automatic Depressurization System (ADS) which provides a means to rapidly depressurize the primary system to a pressure where low-pressure systems can provide makeup for core cooling in the event of a small or medium break Loss of Coolant Accident.

No actions to reduce the frequency or consequence are necessary.

REPORT ABILITY Since ADS is a single train system, PNPS currently believes that this event is reportable under 10 CFR 50.73(a)(2)(i)(B), as a condition prohibited by Technical Specifications and also, potentially reportable under 50.73(a)(2)(v)(B) and 50.73(a)(2)(v)(D), a condition that could have prevented fulfillment of a safet~

function needed to remove residual heat and mitigate the consequences of an accident.

PREVIOUS EVENTS LER 2015-002-00, Main Steam Safety Relief Valves Determined to be Inoperable Following Evaluation LER 2013-002-00 and -01, SRV-38 Safety Relief Valve Declared Inoperable Due to Leakage and Setpoint Drift LER 2011-007-00, Safety Relief Valve Declared Inoperable Due to Leakage REFERENCES CR-PNP-2017-5067 CR-PNP-2017-5386 CR-PNP-2017-6183 Page 3 of 3