NL-20-0230, Edwin I. Hatch Nuclear Plant, Unit 2, Primary Containment Penetration Exceeded Maximum Allowable Primary Containment Leakage Rate (La)

Edwin I. Hatch Nuclear Plant, Unit 2, Primary Containment Penetration Exceeded Maximum Allowable Primary Containment Leakage Rate (La)

ML20062F391
Person / Time
Site:	Hatch
Issue date:	03/02/2020
From:	Dean E Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-20-0230 LER 2020-001-00
Download: ML20062F391 (5)

Text

A Southern Nuclear E. D. Dean Ill Vice President - Plant Hatch l'hulll dwin I Hld<h I IU:!X Hat'h l'arkw<ty 1\nnh llaxlcy, GA 1 1513

~ 12 517 ~xw 1ul 912 11\1\ 2077 fax MAR 0 2 2020 Docket Nos.: 50-366 NL-20-0230 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Unit 2 Primary Containment Penetration Exceeded Maximum Allowable Primary Containment Leakage Rate (La)

Ladies and Gentlemen:

In accordance with the requirements of 10 CFR 50.73 (a)(2)(i)(B). 10 CFA 50.73 (a)(2)(ii)(A), and 10 CFR 50.73 (a)(2)(v)(C). Southem Nuclear Operating Company hereby submits the enclosed Licensee Event Report.

This letter contains no NRC commitments. If you have any questions, please contact the Hatch Licensing Manager, Jimmy Collins at 912.537.2342 Respectfully submitted, r!lfl~

E. D. Dean Vice President - Hatch EDD/JEUSCM

Enclosure:

LEA 2020-001 -00 Cc: Regional Administrator, Region II NRR Project Manager- Hatch Senior Resident Inspector- Hatch RTYPE: CHA02.004

Edwin I. Hatch Nuclear Plant Unit 2 Licensee Event Report 2020-001-00 Primary Containment Penetration Exceeded Maximum Allowable Primary Containment Leakage Rate (La)

Enclosure LER 2020-001-00

NRC FORM366 (04-2018)

U.S. NUCLEAR REGULATORY COMMISSION

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1. Facility Name 2. Docket Number 3.Page Edwin I. Hatch Nuclear Plant Unit 2 05000 366 1 OF 3

4. nue Primary Containment Penetration Exceeded Maximum Allowable Primary Containment Leakage Rate (La)

5. Event Date 6. LER Number 7. Report Date B. Other Facilities Involved I I Sequontlal Rev oclllty Name Docket Number Month O.y Year Year Monll\ Day Year Numb.,. 05000 No.

Docket Number 3 :l.

Foclllty Nome 01 04 2020 2020 001 00 21111..() 05000

9. Operating Mode 11. This Report Is Submitted Pursuant to the Requirements of 10 CFR §: (Check all that apply) 0 20.2201(b) 0 20.220J(a)(3)(1) 0 50 .73(a)(2)(ii)(A) 0 S0.73(a)(2}(Yiii)(A) 0 20.2201(d) 0 20.2203(a)(3)(ii)

D 50 73(a)(2)(ii)(B}

D 50.73(a)(2}(viij)(B) 1 D 20.2203(a)(1) 0 20.2203(a)(4)

D 50 73(a)I 2)(iii)

D 50.7J(a)(2)(1x)(A) 0 20.2203(a}(2)(i) 0 50.36(c)(1)(1)(A)

D sa.73(aJ(2)(iv)(A) D 50,73(a)(2)(x)

10. Power Level 0 20.2203(a)(2)(ii} 0 50.36(c)(1)(1 )(A) D 50,73(a](2)(v)(A) D 73.71(a)(4) 0 20.2203(a}(2)(ill) 0 50.36(c)(2)

D S0.73(a)(2)(v)(B) 0 73.71(a)(5) 0 20 220J(a)(2}(iv) D 50.46(8)(3)(11) 0 50,73(a)(2)(v)(C) 0 73.77(8)(1) 100 0 20 2203(a}(2)(v) D 50.73(a)(2)(1)(A) D 50.73(a)(2)(v)(D) 0 73.n(a)(2)(1)

D 20.2203(a)(2)(vi) IZ] 50.7J(a)(2)(i)(B)

D 50.73(a)(2)(vii) D 73.77(a)(2)(iQ D 50.73(a)(2)(i)(C)

D Other (Specify In Abstract below or In NRC Form 366A)

12. licensee Contact forthls LER licensee Contact Telephone Number (Include Area Code)

Jimmy Collins- Licensing Manager 912-537-2342 I I

13. Complete One Line for each Component Failure Described In this Report I I I I C.UIO Syatem Manutac1u~ Reportable to ICES Cauao Syalem Component Monulac1um Reportable to ICES X BB Cois"v""'

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14. Supplemental Report Expected Monllt Day Year

15. Expected 0

Submission Date v.. (If yes, complete 15. Elcpected Submission Date) [Z] No Abstract (Limit to 1400 spaces, I.e., approximately 14 single-spaced typewritten lines)

On January 4, 2020 at 1109 EST, with Unit 2 operating at 100% rated thermal power, it was determined that the maximum allowable primary containment leakage rate (La) as defined in 10CFR50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors", had been exceeded under postulated accident conditions .

Engineering troubleshooting efforts identified the degraded primary containment penetration and noted that leakage past two primary containment isolation valves (PCIVs) was causing La to be exceeded under postulated accident conditions. Additional valves downstream of the PCIVs were closed to retum primary containment back to operable status.

The cause of the PC IV failures is currently unknown and will be determined during valve disassembly during a planned outage.

At that time, the PCIVs will be repaired and returned to operable status.

NRC FORM 366 (04*2018)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150..0104 EXPIRES: 03/31/2020 (04-2019)

Estimated burden per respon~e Ill complr with this manda!oly colledion reque$t 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Reported LICENSEE EVENT REPORT (LER) lessons learned arv incorporated inlo 1he icensing proce$$ and fed bad< 1o lndu$1ry, Send comments regatding burden estima18 Ill lhe Information Setvlces Branch (T*2 F43), U. S Nudeilf RegWiory CONTINUATION SHEET Commission WeshingiOn, DC 20555-0001, or by &-mai to lnlocollects.Resource@nrc.gov and to lhe Oe$k Offocer, 011\ca ollnlonnation and Regulatory Allairs, NEOB-10202. (J15G-OIG4). Office ol Management ard Budge~ Washington. DC 20503. H a mean$ used 10 ~~~~ an InformatiOn (See NUREG-1022, R.3 for lnstrucUon and guidance for completing this form co8ectlon does not display a currenlfy valid OMB control number, lhe NRC may not conduct Of htto*ttwww,nrc,govtreadlng*rm/doc-collections/nuregs/staff/srl022/Q/\ $p0nsor, and a pe~n Is not required to respord IO, lhe lnlonnatiOn cotJection.

1. FACIUTY NAME 2. DOCKET NUMBER 3. LER NUMBER c:JYEAR SEQUENTlAL REV 1- 0 NUMBER NO.

05000-1 I L:J -I Edwin I. Hatch Nuclear Plant Unit 2 366 001 NARRAllVE EVENT DESCRIPTION On January 4, 2020, at 1109 EST, with Unit 2 operating at 100% rated thermal power, it was determined that the maximum allowable primary containment leakage rate (La) as specified in the Technical Specifications and defined in 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors", had been exceeded under postulated accident conditions.

Engineering troubleshooting efforts identified that leakage past the drywell ventilation penetration inboard and outboard isolation valves was causing La to be exceeded under postulated accident conditions. This represented a failure of the referenced penetration to maintain primary containment integrity. Compensatory measures were Implemented to close and deactivate isolation valves downstream of the Primary Containment Isolation valves (PCIVs) (ISV) to isolate the leaking penetration and return primary containment to operable status. The downstream valves are safety-related and constructed to ASME Section Ill Class 2 with satisfactory Local Leak Rate Tests (LLRTs).

Failed Components Information:

Master Parts List Number: 2T48F319 and 2T48F320 Manufacturer: Fisher Controls Company Model Number: Model 9220 Type: Isolation Valves This event is reportable per 10 CFR 50.73(a)(2)(ii)(A) due to one of the plant's principle safety barriers being seriously degraded. This event is also reportable per 10 CFR 50.73(a)(2)(v)(C) for an event or condition that could have prevented fulfillment of a safety function that is needed to control the release of radioactive material. This event has been classified as a Safety System Functional Failure under NEI 99-02. This event is also reportable per 10 CFR 50.73(a)(2)(i)(B) for a condition prohibited by Technical Specifications because primary containment was inoperable in excess of the allotted Limited Condition of Operation (LCO) timeframe.

EVENT CAUSE ANALYSIS The cause of the PC IV failures cannot be confirmed until valve disassembly. During a planned outage, the PCIVs will be repaired and returned to operable status.

ASSESSMENT OF SAFETY CONSEQUENCES There was no radioactive release to the public, so there were no actual safety consequences as a result of this event.

Upon identification of the excessive leakage rate, the operating crew responded correctly by isolating the leakage path.

The applicable Technical Specifications were properly entered. and required actions were taken within the specified completion time.

The leakage through the degraded primary containment penetration for the as-found condition was determined to exceed La under postulated accident conditions. However, the primary containment leakage rate was below the level required to significantly impact the Core Damage Frequency and Large Early Release Frequency and was of low safety consequence. Additionally, any leakage past the degraded PCIVs during an actual event would have been filtered through SBGT and released through the main stack at an elevated level.

NRC FORM 3661\ (04-2018}

Page 2 of 3

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150..0104 EXPIRES: 03/31/2020 (04-2018)

Estimated burden per response b comply 1\flh lhls mandatory c:olection request 80 hour9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.;. Reported lessons leill!led atelnoo!pOtaled Into lhe &censing process and led back to Industry. Send ccmmeniS

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1. FACIUTY NAME 2. DOCKET NUMBER 3. LER NUMBER Edwin I. Hatch Nuclear Plant Unit 2 05000-1 366 1::::1 YEAR SEQUENTIAL NUMBER REV NO.

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NARRATIVE CORRECTIVE ACTIONS During an upcoming planned outage, the referenced PCIVs will be repaired and returned to operable status.

PREVIOUS SIMILAR ISSUES On February 7, 2017, with Unit 2 in a refueling outage, the same drywell ventilation penetration inboard isolation valve failed LLRT. On February 19, 2017, while still in the refueling outage, the same drywell ventilation penetration outboard isolation valve failed LLRT. This condition represented a failure of the associated penetration to maintain primary containment integrity due to both PCIVs In this penetration flow path exceeding La.

The cause of the PCIVs exceeding La was attributed to inadequate conditions related to the disc sealing ring that was found on both valves. Corrective actions included replacing the ring assemblies and adjusting the set screws on both PCIVs. A satisfactory LLRT was subsequently performed for both valves.

NRC FORM 366A (04-2018) Page 3 of 3