05000336/LER-2015-002, Regarding Degraded Emergency Core Cooling System Check Valve

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Regarding Degraded Emergency Core Cooling System Check Valve
ML15229A166
Person / Time
Site: Millstone Dominion icon.png
Issue date: 08/10/2015
From: Adams M
Dominion, Dominion Nuclear Connecticut
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
15-403 LER 15-002-00
Download: ML15229A166 (8)


LER-2015-002, Regarding Degraded Emergency Core Cooling System Check Valve
Event date:
Report date:
Reporting criterion: 10 CFR 50.73(a)(2)(v), Loss of Safety Function

10 CFR 50.73(a)(2)(ii)

10 CFR 50.73(a)(2)(viii)

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)(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)(i)(A), Completion of TS Shutdown

10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3362015002R00 - NRC Website

text

~Dominion Dominion Nuclear Connecticut, Inc.

Rope Ferry Rd., Waterford, CT 06385 Mailing Address: P.O. Box 128 Warerford, CT 06385 domecom Afh10 20l15 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Serial No.

MPS Lic/LES Docket No.

License No.15-403 R0 50-336 DPR-65 DOMINION NUCLEAR CONNECTICUT. INC.

MILLSTONE POWER STATION UNIT 2 LICENSEE EVENT REPORT 201 5-002-00 DEGRADED EMERGENCY CORE COOLING SYSTEM CHECK VALVE This letter forwards Licensee Event Report (LER) 2015-002-00 documenting an event at Millstone Power Station Unit 2 on June 11, 2015. This LER is being submitted pursuant to 10 CFR 50.73(a)(2)(v)(C). A supplemental LER is planned to be submitted by December 11,2015.

If you have any questions or require additional information, please contact Mr. Thomas G.

Cleary at (860) 447-1791 x3232.

Sincerely, Plant Manager - Millstone Attachments: I Commitments made in this letter: None f695-

Serial No.15-403 Docket No. 50-336 Licensee Event Report 2015-002-00 Page 2 of 2 cc:

U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 R. V. Guzman NRC Project Manager Millstone Units 2 and 3 U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 C-2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station

Serial No.15-403 Docket No. 50-336 Licensee Event Report 2015-002-00 ATTACHMENT LICENSEE EVENT REPORT 2015-002-00 DEGRADED EMERGENCY CORE COOLING SYSTEM CHECK VALVE MILLSTONE POWER STATION UNIT 2 DOMINION NUCLEAR CONNECTICUT, INC.

'NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 01/31/2017

,(02-2014)

  • ,.,.,"",oEstimated burden per reeponse to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.

,,-*'t*_.**-Reported lessons learned are incorporated into the licensing process and fed back to industry.

,~,

Send comments regarding burden estimate to the FOIA, Privacy and Information Collections

°... LICENSEE EVIENT REPORT (LER)

Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by itmte-mail to lnfocollects.Resource @nrc.gov, and to the Desk Ottficer, Ottice of Information and (See Page 2 for required number of RegulatoryAffairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid 0MB digit.S/Ch aracters fo uac lock) control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

[3. PAGE Millstone Power Station Unit 2 05000336J 1 OF 5

4. TITLE Degraded Emergency Core Cooling System Check Valve
5. EVENT DATE
6. LER NUMBER

[

7. REPORT DATE
8. OTHER FACILITIES INVOLVED MOT A EA EUNIA E((FACILITY NAME DOCKET NUMBER MOT A ER YEAR SEUMENILR EVO MONTH DAY YEAR 05000 jFACILITY NAME DOCKET NUMBER 06 11 2015 2015 -

002

- 00 [08 10 2015 05000
9. OPERATING MODE
11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)

[]

20.2201(b)

LII 20.2203(a)(3)(ii)

[]

50.73(a)(2)(ii)(C)

[]

50.73(a)(2)(viii)(

D] 20.2203(a)(1)

LI 20.2203(a)(4)

LI 50.73(a)(2)(ii)(B)

LI 50.73(a)(2)(viii)(B)

L 20.2203(a)(2)(i)

[]

50.36(0)(1)(i)(A)

LI 50.73(a)(2)(iii)

LI 50.73(a)(2)(ix)(A)

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

LI 50.36(c)(1)(ii)(A)

LI 50.73(a)(2)(iv)(A)

LI 50.73(a)(2)(x)

LI 20.2203(a)(2)(iii)

LI 50.36(c)(2)

LI 50.73(a)(2)(v)(A)

LI 73.71(a)(4)

LI 20.2203(a)(2)(fv)

LI 50.46(a)(3)(ii)

LI 50.73(a)(2)(v)(B)

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

LI 50.73(a)(2)(i)(A)

Z] 50.73(a)(2)(v)(C)

LI OTHER

,Lf-I 20.2203(a)(2)(vi)

,LF"I 50.73(a)(2)(i)(B)

L 1-I 50.73(a)(2)(v)(D) specify in Abstract below or in

1. EVENT DESCRIPTION

On June 11, 2015, while Millstone Power Station Unit 2 (MPS2) was in MODE 1 operating at 100 percent power, Engineering identified that due to a degraded check valve, the post-accident radioactivity release rates assumed in the FSAR could be affected. While performing 'B' High Pressure Safety Injection (HPSI) pump in-service testing, the measured flow was lower than expected. Because both trains of HPSI, Low Pressure Safety Injection (LPSI) pumps, and Containment Spray (CS) pumps share a common minimum flow recirculation line back to the Refueling Water Storage Tank (RWST), back-leakage through one of the idle pumps recirculation check valves was postulated as the cause of the observed drop in flow. Troubleshooting was performed, and it was determined that a degraded minimum flow check valve associated with the 'A' CS pump (2-CS-6A) was back-leaking into the "A" train suction line and flowing to the RWST through 2-CS-14A (RWST suction check valve) and normally open 2-CS-13.IA (RWST suction line isolation).

The associated minimum flow isolation valve for 2-CS-6A was closed to eliminate the path and the valve was repaired. 2-CS-14A permitted back flow because the conditions during the test provided insufficient back flow to adequately seat it.

Subsequently, engineering evaluation determined that during a loss of power to one facility or train of the Emergency Core Cooling System (ECCS), this back-leakage flow path could result in more leakage to the RWST than had been previously considered in the accident analysis. This leakage has the potential to adversely affect calculated post-Loss of Coolant Accident recirculation phase radioactivity release rates under some postulated scenarios. Specifically, without power on the affected (i.e.: back-leaking) train, the CS discharge flowpath would not automatically open resulting in leakage to the RWST.

Prior to this discovery, the last time that the 'A' containment spray pump was run was on April 15, 2015, which would have opened 2-CS-6A. On this basis, it was concluded that this condition existed from April 15, 2015 to June 11, 2015. During this period, the 'A' Emergency Diesel Generator (EDG) was inoperable 4 times for a total of approximately 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />. However, all 4 times were for surveillances. At no time during this period was any maintenance done on the 'A' EDG. It is judged that, between accident initiation and commencement of sump recirculation (approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with only one train available), the EDG could have been made available.

This event is being reported as an event or condition that could have prevented fulfillment of a safety function to control the release of radioactive material under 10 CFR 50.73(a)(2)(v)(C). Note: The initial non-emergency report (#51149) of this issue on June 11, 2015 was subsequently updated on July 10, 2015. An additional unidentified release path from the original design of the plant was reported to the NRC as a follow-up notification on July 10, 2015.

BACKGROUND The MPS2 Containment Spray (CS) system functions as an engineered safety feature to limit containment pressure and temperature after a loss-of-coolant accident (LOCA) and'Main Steam Line Break (MSLB) accident, and thereby reduce the potential for leakage of airborne radioactivity to the outside environment. A minimum flow recirculation line is included in the design for recirculating water from the outlet of the pump to the RWST. With the CS system operating during a design basis accident, a small portion of the CS pump discharge flow recirculates to the RWST during the injection

phase; however, the recirculation line is isolated from the RWST when transferring to sump recirculation. All seven ECCSICS pumps (2 LPSl, 2 CS, and 3 H PSI) have minimum flow recirculation lines that tie into one common header to the RWST. Exhibit A attached to this LER is a sketch depicting the configuration.

Evaluation of this failure identified that during a Small Break Loss of Coolant Accident (SBLOCA),

concurrent with a loss of power to the "A" train associated with the leaking 2-CS-6A, the operating (opposite) train HPSI pump would pressurize the common recirculation header during the recirculation phase of the accident as was described above. In the operator response to this postulated event when the 2-CS-6A leaks to the RWST, suction 2-CS-I13.1A is manually isolated terminating the release to the RWST. However, it was additionally identified that the "A" train suction header would pressurize because of the continued back leakage and none of the "A" train pumps flowing due to loss of power and no open flow path (to containment spray or RWST). This pressurization would continue, exceeding the design pressure of the suction header (60 psig) ultimately reaching 500 psig, the lift setpoint for the "A" train shutdown cooling heat exchanger relief valve downstream of the "A" CS pump and the relief valve on the common LPSl discharge header.

These relief valves discharge to the EDST (Emergency Drain Sump Tank) which is located and vented outside the filtered ventilation boundary. Upon emptying of the RWST and initiation of sump recirculation, the described back-leakage flow path would contain sump fluid. This additional unidentified release path from the original design of the plant was reported to the NRC as a follow-up notification on July 10, 2015.

For this condition (2-CS-6A back-leakage) to result in significant radiological consequences, the following accident sequence must occur:

2-CS-6A back-leakage SBLOCA Loss of Offsite Power (LOOP)

Loss of the "A" train of onsite electrical power SBLOCA break size that results in RCS re-pressurization and sustained RCS pressures above 500 psig, such that HPSl would continue to pressurize the recirculation header above the relief valve setpoints Fuel damage While the discussion above relates to 2-CS-6A and the "A" train, back-leakage from any of the minimum flow check valves and a concurrent loss of power to that train would have a similar outcome.

2. CAUSE

The cause of the event was a failed-open check valve. Further analysis indicates a leaking check valve could also cause the problem described in the Assessment of Safety Consequences.

Leakage through these check valves was not considered in the FSAR.

3. ASSESSMENT OF SAFETY CONSEQUENCES

A preliminary assessment reveals that if a recirculation line check valve fails open or leaks, the following conditions could occur:

1. An additional recirculation fiow path exists for the running pumps back to the non-running pumps' respective suction lines via the common minimum flow recirculation line.
2. During a loss of power to one facility or train of the ECCS/CS systems, more leakage to the RWST than had been previously analyzed could result.
3. Under certain small break LOCA conditions, ECCS/CS non-operating train suction piping could be pressurized significantly above its design pressure, and relief valves could lift, sending sump recirculation water to the Equipment Drains Sump Tank (EDST), and result in an unfiltered release of radioactivity, and associated reduction in sump inventory.
4. Low pressure piping portions of ECCS systems could be caused to yield for the failed open check valve case, but not for the leaking check valve case, although would be unlikely to rupture.

This scenario must be evaluated using modified methods. Millstone typically evaluates design basis radiological consequences based on a LBLOCA and associated fuel damage and a SBLOCA model does not exist. Therefore, a preliminary sensitivity model has been developed using best-estimate assumptions (i.e., a gap release source term, 1% iodine partitioning in the EDST, and measured control room filter efficiency and unfiltered leakage) for application in the small break LOCA scenario. Preliminary results show that control room operator dose would remain within the regulatory limits of 10 CFR 50.67 if the release from 2-CS-6A is terminated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and offsite doses would also remain within the regulatory limits of 10 CFR 50.67 (25 REM TEDE offsite dose and 5 REM TEDE Control Room dose) under these assumptions.

The evaluation of the dose consequences is continuing for this event, and a supplement to the LER is planned once the evaluation is completed.

4. CORRECTIVE ACTION

The failed check valve 2-CS-6A has been repaired. Additional corrective actions are being taken in accordance with the station's corrective action program.

5. PREVIOUS OCCURRENCES

None

6. Energy Industry Identification System (EllS) codes:

Pump-P

  • Valve-V
  • Tank-T
  • Safety Injection - SIU.S. NUCLEAR REGULATORY COMMISSION (02-2014)

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET

1. FACILITY NAME [2. DOCKET
6. LER NUMBER j
3. PAGE ISEQUENTIAL REV IYEAR NUMBER NO.

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