05000366/LER-2009-002, Primary Containment Allowable Bypass Leakage Rate Exceeded Due to Failure of Feedwater Check Valve(S)

Primary Containment Allowable Bypass Leakage Rate Exceeded Due to Failure of Feedwater Check Valve(S)

ML091240255
Person / Time
Site:	Hatch
Issue date:	05/04/2009
From:	Madison D Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-09-0691 LER 09-002-00
Download: ML091240255 (5)

LER-2009-002, Primary Containment Allowable Bypass Leakage Rate Exceeded Due to Failure of Feedwater Check Valve(S)

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(viii)(A) 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)(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)(v), Loss of Safety Function
3662009002R00 - NRC Website

text

Dennis R. Madison Southern Nuclear Vice President - Hatch Operating Company. Inc.

Plant Edwin I Hatch 11 aZ8 Hatch Parkway North Baxley, Georgia 31513 Tel 912537.5859 Fax 912.3662077 SOUTHERN A May 4,2009 COMPANY Docket No.:

50-366 NL-09-0691 U. S. Nuclear Regulatory Commission AnN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Licensee Event Report Primary Containment Allowable Bypass Leakage Rate Exceeded due to Failure of Feedwater Check Valve(s)

Ladies and Gentlemen:

In accordance with the requirements of 10 CFR 50.73(a)(2)(i)(B), and 10 CFR 50.73(a)(2)(ii)(A) Southern Nuclear Operating Company is submitting the enclosed Licensee Event Report (LER) concerning primary containment allowable bypass leakage rate exceeded due to failure of Feedwater check valve(s).

This letter contains no NRC commitments. If you have any questions, please advise.

Sincerely,

/O~'7??~

D. R. Madison Vice President - Hatch DRM/MJKldaj Enclosure: LER 2-2009-002 cc: Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Ms. P. M. Marino, Vice President - Engineering RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Hatch Mr. J. A. Hickey, Senior Resident Inspector - Hatch

APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08f3112010 9-2007)

"RC FORM 388 U.S. NUCLEAR REGULATORY COMMISSION Estimated burden per response to comply with this mandatory collection request: 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />. Reported lessons leamed are incorporated into the licensing process and fed back to industry. Send comments re~arding burden estimate to the Records and FOIAIPrivacy Service Branch

- 5 F52), U.S.

Nuclear Re?culatOry Commission, Washington, DC 20555'()()()1, or ~ intemet LICENSEE EVENT REPORT (LER) e-man to In ocollects@nrc80v, and to the Desk Officer. Office of In ormation and Regulatory Affairs. NE B-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currenliy valid OMB control numbar, lhe NRC may not conduct or sponsor. and a person is not required to respond to. the inlormation colleciion.

2. DOCKET NUMBER

1. FACILITY NAME

\\3. PAGE 05000366 1 OF 4 Edwin 1. Hatch Nuclear Plant Unit 2

4. nTLE Primary Containment Allowable Bypass Leakage Rate Exceeded Due To Failure of Feedwater Check Valve(s)

8. OTHER FACILInES INVOLVED FACILITY NAME

7. REPORT DATE

6. LER NUMBER

5. EVENT DATE DOCKET NUt.AIlER SEQUENTIAL REV MONTH DAY YEAR YEAR MONTH DAY YEAR 05000 NO.

NUMBER DOCKET NUMBER FACILITY NAME 2009 002 0

05 2009 2009 04 13 03 05000

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

9. OPERAnNG MODE o 20.2201 (b) o 20.2203(a)(3)(i) o SO.73(a)(2)(i)(C) o 50.73(a)(2)(vii) 5 o 20.2201 (d) o 20.2203(a)(3)(ii) 181 50.73(a)(2)(ii)(A) o 50.73(a)(2)(viii)(A) o 20.2203(a)(1) o 20.2203(a)(4) o 50.73(a)(2)(ii)(B) o 50.73(a)(2)(viii)(B) o 20.2203(a)(2)(i) o 50.36(C)(1)(i)(A) o 50.73(a)(2)(iii) o SO.73(a)(2)(ix)(A) o 20.2203(a)(2)(ii) o 50.36(c)(1 )(ii)(A) o 50.73(a)(2)(iv)(A) o 50.73(a)(2)(x)

10. POWER LEVEL o 20.2203(a)(2)(iii) o 50.36(c)(2) o 50.73(a)(2)(v)(A) o 73.71(a)(4) o 20.2203(a)(2)(iv) o 50.46(a)(3)(ii) o 50.73(a)(2)(v)(B) o 73.71(a)(5) 0.00 o 20.2203(a)(2)(v) o 50.73(a)(2)(i)(A) o 50.73(a)(2)(v)(C) o OTHER o 20.2203(a)(2)(vi) 181 50.73(a)(2)(i)(B) o 50.73(a)(2)(v)(D)

Specify in Abstract below or in NRC Form 366A

12. LICENSEE CONTACT FOR THIS LER FACILITY NAME I~ELEPHONE NUMBER (Indude Area Code)

Edwin 1. Hatch I Kathy Underwood, Performance Improvement Supervisor 912-537-5931 MANU REPORTABLE MANU REPORTABLE SYSTEM COMPONENT

CAUSE

SYSTEM COMPONENT

CAUSE

FACTURER TO EPIX FACTURER TO EPIX SHY R344 YES X

SJ SHY X

SJ A585 YES

14. SUPPLEMENTAL REPORT EXPECTED

15. EXPECTED SUBMISSION MONTH DAY YEAR o YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 181 NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)

On March 13,2009, Unit 2 was in the Refuel mode with fuel in the reactor vessel. Personnel were evaluating local leak rate testing (LLRT) on the valves and penetrations which comprise the primary containment bypass leakage isolation valves. The valves are classified as secondary containment bypass valves since the piping is routed outside the secondary containment boundary. It was determined that the "A" feedwater line, which is considered to be a water filled line, had a tested leakage value that would result in the line losing its water seal prior to thirty days after the initiation of a design basis accident, and thus the leakage should be considered as part of the bypass leakage. As part of the bypass leakage the measured value when converted to an equivalent air leakage value exceeds the allowable bypass leakage criteria.

The cause of the feedwater valve test failures was misalignment caused by internal wear and missing bearing cover lock pins for the outboard valve and misalignment caused by excessive clearance between the hinge pin and the disc, and the hinge pin adjustment had changed over the operating cycle.

Corrective actions for this event included rebuilding the valves with new parts. replacing the missing parts, and performing a modification to account for the incorrect hinge pin length.

PRINTED ON RECYCLED PAPER NRC FORM 366 (9-2007) U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

(9-2007)

CONTINUATION SHEET

6. LER NUMBER

3. PAGE

2. DOCKET

1. FACILITY NAME SEQUENTIAL IREVISION YEAR I

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Edwin I. Hatch Nuclear Plant Unit 2 05000366 2009 002 o

NARRATlVE (If more space is required, use additional copies of NRC Form 366A)

PLANT AND SYSTEM IDENTIFICATION

General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EllS Code XX).

DESCRIPTION OF EVENT

On March 13, 2009, Unit 2 was in the Refuel mode with fuel in the reactor vessel. Personnel were evaluating local leak rate testing (LLRT) on the valves and penetrations which comprise the Primary Containment (EllS Code NH) bypass leakage isolation valves. These valves are classified as Secondary Containment (EllS Code NG) bypass valves since the piping is routed outside the secondary containment boundary. It was detennined that the "A" feedwater line, which is assumed to be a water filled line for 30 days after the initiation of a design basis accident (DBA), had a tested leakage value that would result in the line losing its water seal prior to the thirty day time frame following initiation of a DBA. The piping has a path which exits the secondary containment boundary. Therefore, the leakage from this penetration would become a source of bypass leakage when the water seal is lost. This LLRT is performed using water and the leakage criteria assume the leakage through the penetration will be water. When a comparable air leakage rate is calculated using the primary containment pressure at the time the water seal is lost, the leakage rate would exceed the allowed bypass leakage of 0.009 La (L sub a). This penetration is not included in the list of bypass leakage penetrations since the line is nonnally filled with water. However, recognizing that the line exits secondary containment and presents a potential leakage path if the line loses its water seal, it is appropriate to consider this line as a bypass leakage line when the valve leakage is above the allowable limit.

As a reSUlt, the above condition exceeds the 0.009 La (L sub a) bypass leakage limit.

During the investigation of this event the basis for the LLRT leakage acceptance criteria for penetrations 2T23-XOO9A and 2T23-XOO9B which includes 2B21-FOlOA, 2B21-FOlOB, 2B21-F077A, and 2B21-F077B was re-evaluated and determined to require revision to a lower allowable leakage limit. When reviewing previous outage test data, where this updated allowable leakage limit would have been applicable, it was detennined that in March of 2007 the "as left" leakage for 2B21-FOlOB was left greater than the updated leakage limit. As a result the tested leakage through this one valve was above the allowed leakage limit to ensure compliance with the 0.009 La (L sub a) leakage criteria of the Technical Specifications. The "as left" leakage value for the companion valve (2B21-F077B) in this penetration was below the allowable leakage limit. Therefore the overall penetration leakage was within the allowable leakage limit. 10 CFR 50 Appendix J, Option B requires that when starting the reactor from a refueling outage the maximum pathway leakage of the primary containment penetrations must not exceed the established limit. With the one valve exceeding the updated allowable leakage limit the criteria was not met.

CAUSE OF EVENT

The cause of the feedwater outboard valve test failure (only the "A" line was affected) was misalignment caused by internal wear, and missing bearing cover lock pins (this is valve 2B21-F077A).

The cause of the feedwater inboard valve test failures (the "A" and "B" lines were affected) was misalignment caused by excessive clearance between the hinge pin and the disc and, the hinge pin adjustment had changed over the operating cycle.

PRINTED ON RECYCLED PAPER NRC FORM 368A (~2007) (9-2007)

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION

1. FACILITY NAME

2. DOCKET

6. LER NUMBER

3. PAGE Edwin I. Hatch Nuclear Plant Unit 2 05000366 YEAR I SEQUENTIAL IREVISION NUMBER NUMBER 3

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2009 002 0

The cause of the failure to meet the technical specification requirement of 0.009 La (L sub a) for reactor start up maximum pathway leakage was an incorrect calculation.

REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT This event is reportable per 10 CFR 50.73 (a)(2)(ii)(A) because an event occurred which resulted in one of the plant's principal safety barriers being degraded. Specifically, the primary containment isolation function involving secondary containment bypass valves was found to not satisfy the leakage requirements of the Technical Specifications. In addition a reportable condition exists in that the reactor was operated with one valve having a leakage value above the allowable leakage rate which results in a condition prohibited by the technical specifications and is reportable per 10 CFR 50.73(a)(2)(i)(B).

The function of the primary containment is to isolate and contain fission products released from the reactor primary system following a design basis accident (DBA) and to confine the postulated release of radioactive material. The primary containment consists of a steel vessel which surrounds the reactor primary system and provides a barrier against the uncontrolled release of radioactive material to the environment. Some leakage from the primary containment is assumed to occur, although the majority of the leakage is assumed to be released into the secondary containment. The total allowable leakage rate for the primary containment is designated La (L sub a) and is equal to 1.2 percent by weight of the contained air volume per day, most of which is assumed to occur within the secondary containment where it would be treated by the Standby Gas Treatment system (SBGT) (EllS Code BH) before being released at an elevated point through the Main Stack (EllS Code VL). However, some small amount of leakage is assumed to occur outside secondary containment where it is released without being treated by the SBGT system. Valves located in primary containment penetrations whose pipes lead outside the secondary containment are potential sources of such untreated leakage, so these valves are termed "secondary containment bypass valves". Since the atmospheres in such areas would not be filtered by the SBGT system, the allowable leakage through these valves is specifically addressed by the Technical Specifications, and is limited to a total of 0.009 of La (L sub a). The leakage rates measured in this event were greater than this amount.

The Final Safety Analysis Report (FSAR) for Plant Hatch Unit 2 designates the DBA as the break of a Reactor Recirculation system (EllS Code AD) pipe which results in the rapid depressurization of the reactor vessel to the primary containment. However, the FSAR analysis shows that, for such an accident, resulting peak fuel cladding temperatures would be less than those required to produce damage to the fuel. The plant specific SAFERIGESTR analysis for this accident scenario shows that no damage to the fuel cladding would occur. Therefore, by this analysis, the only radioactive materials present in the released coolant would be those already present due to normal operation and the small additional amount of contaminated or activated crud released from vessel internals and primary system piping during the initial stages of the transient. This analysis applies to all operating conditions.

CORRECTIVE ACTIONS

2B21-F077A, the outboard valve was rebuilt using new parts, and the missing bearing cover locking pins were replaced.

2B21-FOlOA and 2B21-FOlOB were repaired by performing a modification to account for the excessive clearance between the hinge pin and the disc and to prevent the hinge pin adjustment from changing.

PRINTED ON RECYCLED PAPER NRC FORM 3e6A (9-2007)

NRC FORM 386A (9-2007)

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION

1. FACILITY NAME

2. DOCKET

6. LER NUMBER

3. PAGE Edwin I. Hatch Nuclear Plant Unit 2 05000366 YEAR I SEQUENTIAL

!REVISION NUMBER NUMBER 4

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The calculation for the allowable leakage limit through these two penetrations has been updated.

ADDITIQNAL INFORMAnON Other Systems Affected: No systems other than those already mentioned in this report were affected by this event.

Failed Components Information

None

Failed Components Information

Master Parts List Number:

2B21-FOlOA & FOlOB Manufacturer: Rockwell International Model Number: 970 Type: Valve, Shutoff Manufacturer Code: R344 Master Parts List Number:

2B21-F077A Manufacturer: Atwood and Morrill Model Number: CL 900 Type: Valve, Shutoff Manufacturer Code: A585 EllS System Code: SJ Reportable to EPIX: Yes Root Cause Code: X EllS Component Code: SHV EllS System Code: SJ Reportable to EPIX: Yes Root Cause Code: X EllS Component Code: SHY Commitment Information: This report does not create any permanent licensing commitments.

Previous Similar Events

LER 2-2007-002 documents a similar event for the'A' feedwater line where both the inboard and outboard valves failed the LLRT testing. Corrective actions for that event did not take into account the potential impact of not maintaining the internal tolerances and dimensions identified in this event and did not identify the incorrect length of the hinge pin on the inboard valve.

PRINTED ON RECYCLED PAPER