05000366/LER-2011-001, Regarding Primary Containment Isolation Penetration Exceeded Overall Allowable Technical Specification Leakage Limits

Regarding Primary Containment Isolation Penetration Exceeded Overall Allowable Technical Specification Leakage Limits

ML111640419
Person / Time
Site:	Hatch
Issue date:	06/10/2011
From:	Madison D Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NL-11-1038 LER 11-001-00
Download: ML111640419 (8)

LER-2011-001, Regarding Primary Containment Isolation Penetration Exceeded Overall Allowable Technical Specification Leakage Limits

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2) 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)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(8)
3662011001R00 - NRC Website

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OellllIS It Madison

$ouilmrn NiICle.i[

OperatiJlIj Company. inc.

SOUTHERN ",'

COMPANY June 1 0, 2011 Docket Nos.: 50-366 NL-11-1038 U. S. Nuclear Regulatory Commission A TTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Unit 2 Licensee Event Report 2-2011-001 Primary Containment Isolation Penetration Exceeded Overall Allowable Technical Specification Leakage Limits Ladies and Gentlemen:

In accordance with the requirements of 10 CFR 50.73(a)(2)(ii), Southern Nuclear Operating Company (SNC) is submitting the enclosed licensee event report (LER) concerning a primary containment isolation penetration exceedng overall allowable Technical Specification leakage limits.

This fetter contains no NRC commitments. If you have any questions, please contact Doug McKinney at (205) 992-5982.

Respectfully submitted, D. R. Madison Vice President - Hatch DRM/RLG/lac Enclosure: LER 2-2011-001

U. S. Nuclear Regulatory Commission NL-II-I038 Page 2 cc: Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Ms. P. M. Marino, Vice President - Engineering Mr. M. J. Ajluni, Nuclear Licensing Director RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. P. G. Boyle, NRR Project Manager Mr. E. D. Morris, Senior Resident Inspector - Hatch

Enclosure Edwin I. Hatch Nuclear Plant Licensee Event Report 2-2011-001 Primary Containment Isolation Penetration Exceeded Overall Allowable Technical Specification Leakage Limits

~RC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 9-2007)

Estimated burden per response 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 />. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Records and FOIAIPrivacy Service Branch (T-S F53), US.

Nuclear Regulatory CommiSSion, Washington, DC 20555-0001, or by internet LICENSEE EVENT REPORT (LER) e-mail to infocollectuesources@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs. 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 OMB control number. the NRC may nO! conduct or sponsor, and a person is not required to respond to. the information collection.

13. PAGE Edwin I. Hatch Nuclear Plant Unit 2 05000366 1

1 OF 5

4. TITLE Primary Containment Isolation Penetration Exceeded Overall Allowable Technical Specification Leakage Limits

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

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

MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 04 16 20 I I 20 I I - 00 I -

0 06 10 2011 05000

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check a/l that apply) o 20.2201 (b) 0 20.2203(a)(3)(i) 0 50.73(a)(2)(i)(C) 0 50. 73(a)(2)(vii) 5 o 20.2201(d) 0 20.2203(a)(3)(11) 1:81 50.73(a)(2)(ii)(A) 0 50.73(a)(2)(vIII)(A) o 20.2203(a)(1) 0 20.2203(a)(4) 0 50.73(a)(2)(11)(8) 0 50.73(a)(2)(vlli)(8) 1-______-fD 20.2203(a)(2)(I) 0 50.36(c)(1)(i)(A) 0 50.73(a)(2)(lli) 0 50.73(a)(2)(ix)(A)

10. POWER LEVEL 0 20.2203(a)(2)(ii) 0 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A) 0 50.73(a)(2)(x) o 20.2203(a)(2)(ill) 0 50.36(c)(2) 0 50.73(a)(2)(v)(A) 0 73.71(a)(4) o 20.2203(a)(2)(iv) 0 50.46(a)(3)Oi) 0 50,73(a)(2)(v)(8) 0 73.71(a)(5) o o 20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A) 0 50.73(a)(2)(v)(C) 0 OTHER o 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(8) 0 50.73(a)(2)(v)(D)

Specify in Abstract below or in NRC Form 366A

12. LICENSEE CONTACT FOR THIS LER ACILITY NAME 1TELEPHONE NUMBER (Include Are. Code)

Edwin I. Hatch I Steven Tipps Principal Engineer Licensing 1912-537-5880 1-____ __,1_3_.C_O-'--M_P_LE_T,E_O.;..N

.....E'----L_INE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT REPORTABLE

CAUSE

SYSTEM! COMPONENT

CAUSE

SYSTEM \\COMPONENT FA~~~E-:-

RE~g~~~~LE FA~~~~ER TO EPIX y

x I BB SHY FI30

14. SUPPLEMENTAL REPORT EXPECTED

15. EXPECTED SUBMISSION I MONTH I DAY YEAR o YES (If yes, complete 15. EXPECTED SUBMISSION DA TE)

!81 NO DATE ABSTRACT (Limit to 1400 spaces, i. e.. approximately 15 single-spaced typewritten lines)

On April 16,2011, during the Hatch Nuclear Plant outage, an LLRT was performed on torus purge supply primary containment isolation valve (PC IV) 2T48-F324 which is associated with primary containment penetration 2T23 X205. At that time, plant engineers and technicians were perfonning a local leak rate testing (LLRT) for penetration 2T23-X205 when it was discovered that both PCIVs had failed their LLRTs for this penetration, This resulted in the penetration leakage exceeding the overall allowable leakage (La) required by the Tech Specs for primary containment.

The primary cause for the excessive leakage for valve 2T48-F324 was attributed to "over-travel" of the valve disc which reduced the seat contact with the disc. This was caused by less than adequate procedure guidance for adjusting the valve travel. The excessive leakage for 2T48-F309 was attributed to the valve disc failing to "center" on the seat and wear on the actuator and its linkage. The necessary repairs and adjustments were made and the valves subsequently passed their respective LLRTs. which restored penetration 2T23-X205 to within its Tech Spec leakage limits. Additional procedural guidance is being provided to better control "travel" adjustments and in performance of valve inspections should future LLRT failures occur in which the valves fail to meet their respective leakage acceptance criteria.

NRC FORM 366 (9-2007)

PRINTED ON RECYCLED PAPER

5

PLANT AND SYSTEM IDENTIFICATION

General Electric Boiling Water Reactor Energy Industry Identification System codes appear in the text as fNN1.

DESCRIPTION OF EVENT

On April 16.2011. during the Hatch Nuclear Plant 2R21 outage. an LLRT was performed on torus purge supply primary containment isolation valve (PCIV) 2T4S-F324[BBl which is associated with primary containment penetration 2T23-X20S. At that time. plant engineers and technicians were performing a local leak rate test (LLRT) for penetration 2T23-X205. when it was discovered that both PCIVs (2T48-F309 and 2T48-F324)[BBl had failed their LLRTs for this penctration. This resulted in the penetration leakage exceeding the overall allowable leakage (La) requirements in the Technical Specifications for primary containment The inboard PCIV for the torus purge supply system. 2T48-F309 failed to meet its LLRT acceptance criteria for this penetration when it failed to pressurize the penetration on April 5,20II, thereby exceeding the overall allowable leakage (Ln) required by the Technical SpeCifications for primary containment. Following repair of valve 2T48-F309. the penetration was retested on April 16.201 I. at which time outboard PCIV 2T48-F324 was found to be leaking to such a degree that it too failed to meet its LLRT acceptance criteria for this penetration when it failed to pressurize the penetration. This failure coupled with that of 2T48-F309 caused the penetration leakage to exceed the overall allowable leakage (La) required by the Technical Specifications for primary containment. The valves are IS inch Fisher model 9220 buttert1y valves with an air operated Bettis model 733B-SR Robotarm actuator. The valves are wafer style that utilizes an elastomer T-ring seat to form a seal against its disc when closed and are located in the Torus bay #6 area.

CAUSE OF EVENT

Maintenance personnel attributed the cause of the valve leakage for 2T4S-F309 to the fact that the valve disc was not staying centered within the valve body. resulting in a gap between the disc and the seat at the top of the valve. This was caused by workmanship issues in 2009 when neither the valve vendor nor plant Maintenance personnel ensured the valve disc was properly centered as part of the valve refurbishment and setup dUling that outage. The valve adjustments made during the 2009 outage were sufficient to allow the valve to pass its LLRT with an "as leff' leakage of 639 sccm. but the adjustments were made external to the valve which did not ensure the disc was properly centered. There was also inadequate procedural guidance for identifying or replacing worn components for the inboard isolation valve 2T48-F309 body. The necessary adjustments were made to the valve during the 2011 outage which included properly centering the valve disc to ensure the required contact was made between the valve disc and scat and for proper setup of the val ve.

Additional investigation revealed that the cOl1'ective actions for previous failures to 2T48-F309 were less than adequate in repairing the valve to ensure the ongoing reliability of the valve. This valve has failed to previously meet its LLRT acceptance criteria since the refueling outage in 2005, but the penetration minimum pathway leakage was not previously exceeded for this penetration until the current 2011 outage. Based on a review of work packages the replacement of the T-ring and adjustment of valve seat and disc was apparently done to gain a better seal as the corrective action of choice to correct excessive leakage for this valve during these previous outages. Because of recent failures and resulting Maintenance observations during repairs and adjustments. engineering judgment indicates that a contributor to this and previous failures is continuing actuator wear. However, the more detailed work on the valve along with ensuring its proper setup during the 201 I outage should ensure reliable operation during the current operating cycle following the 2011 refueling LICENSEE EVENT REPORT (LER) u.s. NUCLEAR REGULATORY COMMISSION 110-2010)

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outage. Following the repair and adjustments valve 2T48-FJ09 successfully passed its LLRT with an "as left" leakage of 33 sccm against an acceptance criteria of 4.850 sccm.

Similarly. LLRT test pressure could not be reached for valve 2T48-F324 due to the size of the leakage past the valve seat which was determined to be caused by indications found on the valve seat at the 10 and 6 o'clock positions. The valve dise was apparently over-traveling, preventing a good seal from developing between the disc and scat that apparently eaused the damage. During the replacement of the valve T-ling, Maintenance mechanics noticed that the disc moved when the valve's retaining ring / internal stop was removed dUling T ring replacement. This is evidence that the intemal stops of the valve were being utilized for controlling valve travel rather than the actuator stops which allowed the disc to "long travel", Review of the valve's vendor manual and sHe procedures recommend utilization of the actuator stops rather than valve internal stops during its set up. This is done to avoid potentially overstressing internal valve components and potentially causing damage if the actuator is not adjusted appropriately. Additionally. procedure 52PM-MNT-OI I -0, 'Bettis Robotarm Valve Actuator Inspection', sections 5.1.1 and 7.2.5.4 cautions about potential damage that can occur when allowing the valve disc to seat against the inLemal valve stop rather than its actuator stop. That is also why the valve's internal stop is backed off from the disc during valve set up in accordance with procedure 52PM-T48-013-0, 'Purge and Vent Valve T-Ring Replacement'. after closure of the valve with its actuator to allow the actuator stop to govern the travel of valve.

Following the repair and adjustment of2T48-F324 a successful LLRT was subsequently performed with an "as left" leakage of 33 seem against leakage acceptance critelia of 4.850 sccm.

The successful LLRTs performed for 2T48-F309 and 2T48-F324 confirmed that the "as left'* leakage for penetration 2T23-X205 met the leakage acceptance criteria that is considered for input into the total primary containment leakage to ensure the Technical Specification allowable leakage was satisfied for this penetration.

REPORT ABILITY ANALYSIS AND SAFETY ASSESSMENT This event is reportable per 10 CFR 50.73 (a)(2)(ii)(A), because a condition was identified which resulted in the degradation of one of the plant's plincipal safety barriers. Specifically. the primary containment pathway through penetration 2T23-X205, containing torus purge supply system PCIVs 2T48-FJ09 and 2T48-F324 exceeded the allowable leakage established by the plant's Technical Specifications.

The function of the PCIVs. in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) to within limits. Primary containment isolation ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a DBA. The OPERABILITY requirements for PCIVs help ensure that an adequate primary containment boundary is maintained during and after an accident by minimizing potential paths to the environment. Therefore. the OPERABILITY requirements provide assurance that the pIimary containment function assumed in the safety analyses will be maintained. Two barriers in seIies are provided for each penetralion so that no single credible failure or malfunction of an active component can result in a loss of isolation or leakage that exceeds limits assumed in (he safety analyses. The DBAs that result in a release of radioactive material for which the consequences are mitigated by PCIVs are a LOCA and a main steam line break (MSLB). In the analysis for each of these accidents, it is assumed that PCTVs are either closed or close within the required isolation times following event initiation. This ensures that potential paths to the environment through PClVs (including primary containment purge valves) are minimized.

The single failure erileIion required to be imposed in the conduct of unit safety analyses was considered in the original design of the primary containment purge valves. Two valves in seIies on each purge line provide assurance that both the supply and exhaust lines could be isolated even if a single failure occurred. The LICENSEE EVENT REPORT (LER) u.s. NUCLEAR REGULATORY COMMISSION (1 ()"2010)

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3. PAGE SEQUENTIAL IREVISION YEAR NUMBER

• NUMBER Edwin I. Hatch Nuclear Plant Unit 2 05000366 I

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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. PCIVs form a part of the primary containment boundary. The PCIV safety function is related to minimizing the loss of reactor coolant inventory and establishing the plimary containment boundary duting a DBA. Some leakage from the Plimary 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 as La and is equal to 1.2 percent by weight of the contained air volume per day. For Plant Hatch Unit 2, this equates to a total allowable leakage of 60,432 accm or 254,937 sccm, most of which is assumed to occur within the Secondary Containment where it will be treated by the Standby Gas Treatment System (SGT) before being released at an elevated point through the Main Stack.

Even though the total leakage through penetration 2T23-X205 would have exceeded La, the leakage would still be released into the Secondary Containment and would be treated prior to the elevated release to the environment. The Final Safety Analysis Report (FSAR) for Plant Hatch Unit 2 designates the Design Basis Accident (DBA) as the break of a Reactor Recirculation System 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 even if additional failures are postulated. such as failures of certain power supplies and certain low pressure emergency core cooling systems. Therefore. by this analysis, the only radioactive materials present in the released coolant would bc 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. Since the leakage would be released to the Secondary Containment and the radioactive mateIials postulated to be present in the reactor coolant would be those already present due to normal operation with small contJibutions to crud released from the reactor coolant system. the leakage would be treated by the SGT system prior to the elevated release. For this reason the release would be within IOCFR I00 offsite dose limits and would be considered of low safety significance.

CORRECTIVE ACTIONS

The necessary adjustments were made to valve 2T48-F309 to ensure the required contact was made between the valve disc and seat. The valve subsequently passed its LLRT following repairs and adjustments that addressed the identified workmanship issues with the valve that occurred in the 2009 refueling outage. The more detailed work on the valve along with ensuring its proper setup during the 2011 outage should ensure reliable operation during the current operating cycle following the 201) refueling outage. The current plans are to obtain the needed parts and replace the worn parts or entire valve if necessary dUling the next scheduled refueling outage.

The necessary adjustments and repairs were made to 2T48-F324 to correct the valve travel and to ensure good contact between the valve disc and seat. The valve subsequently passed its LLRT. PM procedure 52PM-T48 013-0 will be revised to ensure the needed valve travel adjustments in the future are properly made for 2T48 F324 following a failed LLRT with the valve pulled from the pipeline. A caution similar to 52PM-MNT-OII section 5.1.1 and caution at 7.2.5.4 will also be included regarding the likelihood of valve internal damage if actuator stops adjustments are made inappropriately.

ADDITIONAL INFORMATION

Similar LLRT results of the other PCIVs in the 2T48 system were reviewed to determine if the condition and causes extended to those components. Valve 2T48-F3IO was the only other similar valve that failed its "as found" LLRT during this outage in the same manner in that it would not pressurize. The failure of the 2T 48-F3l 0 valve was attributed to T -ring wear. The leakage contribution could be measured through 5 LICENSEE EVENT REPORT (LER) u.s. NUCLEAR REGULATORY COMMISSION (10-2010)

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the corresponding PCIV (2T48-F328) in this penetration and the "as found" would not have resulted in La being exceeded. No other 2T48 valves failed their "as found" LLRTs during the testing performed in the 2011 refueling outage or exhibited similar symptoms that would indicate the presence of adjustment issues. Valve 2T48-F31O was subsequently repaired with a 0 sccm has leff' leakage.

Unit 1 LLRT results were reviewed for the cOITesponding Unit 1 valves and no penetrations containing the similar IT48 valves failed their respective LLRTs during that outage that would cause La to be exceeded for primary containment. However, one Unit I penetration containing valves IT4S-F31O and IT48-F328A did fail to meet the leakage acceptance criteria for its penetration, but the minimum pathway leakage for the penetration could be measured. The contribution of the leakage from this penetration along with the other minimum pathway leakages for primary containment did not exceed La. The valves were repaired and adjusted and successfully met their respective acceptance criteria in the "as left" LLRT.

Failed Components Information

Master Parts List Number: 2T48-F309/2T4S-F324 EllS System Code: BB Manufacturer: Fisher Controls Repoltable to EPIX: Yes Model Number: IS in Fisher model 9220 buttertly valve Root Cause Code: X Type:

Valve. Shutoff EIIS Component Code: SHV Manufacturer Code: Fl30

Previous Similar Events

One Unit I penetration containing valves I T48-F31O and I T48-F32SA did fail to meet the leakage acceptance criteria for its penetration in the 20 I 0 refueling outage, but the minimum pathway leakage for the penetration could still be measured. The cause for 1 T48F3 10 failing to pressurize was attributed to a different cause than those identified in this LER. There was not a recent previous similar event that addressed the causes for excessive leakage reported in this Unit 2 LER. The contribution of the leakage from this penetration along with the other minimum pathway leakages for primary containment did not exceed La.

LER 2-2009-002 documents a similar event for the'A' feedwater line in which both inboard and outboard isolation valves failed the LLRT testing resulting in failure of the affected penetration for bypass leakage. The main feedwater check valves arc a different type valve and the causes for their LLRT failure were also different from those experienced with the 2T48 valves reported in this LER.

The resulting corrective actions from that LER would have no bearing on the work performed on the 2T48 valves and would not have prevented their LLRT failure.

Commitment Information: This report does not create any new permanent licensing commitments.