ML13165A342
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0611212013 U.S. Nuclear Regulatory Commission Operations Center Event Report Paze I
- This Event has been retracted !!!
- This Event has been retracted !!!
Part 21 (PAR)
Event#
48585 Rep Org: GE HITACHI NUCLEAR ENERGY Notification Date I Time: 12/13/2012 08:41 (EDT)
Supplier: GE HITACHI NUCLEAR ENERGY Event Date I Time: 12/13/2012 (EST)
Last Modification: 06/12/2013 Region:
1 Docket #:
City: WILMINGTON Agreement State:
Yes County:
License #:
State: NC NRC Notified by: DALE E. PORTER Notifications: DANIEL HOLODY R1DO HQ Ops Officer: STEVE SANDIN ALAN BLAMEY R2DO Emergency Class: NON EMERGENCY JAMNES CAMERON R3DO 10 CFR Section:
MARK HAIRE R4DO 21.21 (a)(2)
INTERIM EVAL OF DEVIATION NRR PART 21 EMAIL 60-DAY INTERIM REPORT INVOLVING AN EVALUATION OF A DESIGN CHANGE TO MAGNE-BLAST CIRCUIT BREAKERS The following information was received via fax:
"Summary "GE Hitachi Nuclear Energy (GEH) is investigating the adequacy of a Design Change in AM 4.16-350-2C and AM 4.16-350-2H Magne-Blast Circuit Breakers as a result of a breaker failure at a BWR Licensee.
"GEH has not completed the evaluation of this condition to determine reportability under 10 CFR Part 21 and is therefore issuing this 60-day Interim Notification. GEH will close or issue an update on this matter on or before June 14, 2013. Given the early status of the evaluation, GEH has no recommended actions at this time. This 60-day Interim Notification is issued in accordance with 10 CFR Part 21.21(a)(2), and will be sent to all GE BWR/2-6 plants and all PWRs.
"Recommendation "GEH advises licensees to take no action at this time. However, if licensees wish to determine if the extent of this issue is present at their respective plants, they may identify breaker models AM 4.16-350-2H/2C with a date of manufacture prior to October 6, 1971. Other AM breaker styles do not use the booster cylinder as designed in breaker model AM 4.16-350-2H/2C and are not included in this concern. Therefore the extent is limited to the AM 4.16-350-2H/2C models manufactured prior to October 6, 1971.
"US Plants Potentially Affected 0:T-j 19
06/12/2013 U.S. Nuclear Regulatory Commission Operations Center Event Report Page
- This Event has been retracted I!!
- This Event has been retracted !
Part 21 (PAR)
Event #
48585 "Nine Mile Point 1-2, Fermi 2, Columbia, Grand Gulf, River Bend, FitzPatrick, Pilgrim, Vermont Yankee, Clinton, Dresden 2-3, LaSalle 1-2, Limerick 1-2, Oyster Creek, Peach Bottom 2-3, Quad Cities 1-2, Perry 1, Duane Arnold, Cooper, Susquehanna 1-2, Brunswick 1-2, Hope Creek, Hatch 1 - 2, Browns Ferry 1-3, Monticello, Callaway, Palo Verde 1-3, Calvert Cliffs 1-2, Ginna, Arkansas Nuclear One 1-2, Indian Point 2-3, Millstone 2, Millstone 3, North Anna 1-2, Palisades, Surry 1-2, Waterford 3, Catawba 1-2, Oconee 1-3, McGuire 1-2, Braidwood 1-2, Exelon Byron 1-2, Three Mile Island 1, Beaver Valley 1-2, Davis-Besse, Seabrook, St. Lucie 1-2, Turkey Point 3-4, Point Beach 1-2, DC Cook 1-2, Prairie Island 1-2, Fort Calhoun, Diablo Canyon 1-2, Crystal River 3, Robinson, Shearon Harris, Salem 1, Salem 2, Summer, South Texas Project 1-2, San Onofre 2-3, Farley 1-2, Vogtle 1-2, Sequoyah 1-2, Watts Bar 1, Comanche Peak 1-2, and Wolf Creek.
"If you have an questions, please call me at (910) 819-4491.
Dale E. Porter Safety Evaluation Program Manager GE-Hitachi Nuclear Energy Americas LLC" Hitachi Reference Number: MFN-128 RO
- *
- RETRACTION FROM DALE PORTER TO NESTOR MAKRIS ON 6/12/13 AT 0959 EDT ***
The following information was received via fax:
"[GE Hitachi Nuclear Energy] GEH has completed all testing and evaluations and has determined that the condition previously described in MFN 12-128 RO is not a reportable condition under 10 CFR Part 21. This information will be sent to all GE BWR/2-6 plants and all PWRs that were previously notified under the 10CFR21.21(a)(2) communication."
Notified R1 DO (Dentel), R2DO (Ehrhardt), R3DO (Lara), R4DO (Gepford), and Part 21 Group via email.
'2
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06-12-2013 1 /9 HITACHI GE Hitachi Nuclear Energy To: NRC Operations Center Fax: 301-816-5151 From: Lisa K. Schichlein, on behalf of Fax: 910-819-4829 Dale E. Porter, GE Hitachi Nuclear Energy Date/Time: June 12, 2013 10:00 am Number of pages, including this one: 9 Comments:
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06-12-2013 2/9 HITACHI GE Hitachi Nuclear Energy Dale E. Porter GE-Hitachi Nuclear Energy Americas LLC June 12, 2013 Safety Evaluation Program Manager MFN 12-128 R1 3901 Castle Hayne Rd.
Wilmington, NC 28401 USA U.S. Nuclear Regulatory Commission T 910 819-4491 Attn: Document Control Desk Dale.Porter@GE.Com Washington, DC 20555-0001
Subject:
Update to Part 21 60-Day Interim Report Notification:
Adequacy of Design Change in AM Magne-Blast Circuit Breakers This letter is issued to provide information concerning an evaluation completed by GE Hitachi Nuclear Energy (GEH) regarding the adequacy of a Design Change in Magne-Blast Circuit Breakers with model numbers AM 4.16-350-2C and AM 4.16-350-2H. As stated herein, GEH has completed all testing and evaluations and has determined that the issue raised and documented in MFN 12-128 RO on December 13, 2012 is not a reportable condition in accordance with the requirements of 1 OCFR Part 21.
If you have any questions, please call me at (910) 819-4491.
Sincerely, Dale E. Porter Safety Evaluation Program Manager GE-Hitachi Nuclear Energy Americas LLC Attachments:
- 1. Description of Evaluation
- 2.
US Plants Previously Notified
References:
- 1. MFN 12-128 RO, Part 21 60-Day Interim Report Notification: Adequacy of Design Change in AM Magne-Blast Circuit Breakers, Dated: December 13, 2012.
cc:
S. S. Phiipott, USNRC S. J. Pannier, USNRC
- 0. Tabatabai-Yazdi, USNRC
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06-12-2013 3/9 MFN 12-128 R1 Page 2 of 2 J. G. Head, GEH G. A. Watford, GEH P. L. Campbell, GEH Washington PRC File DRF Section No. 0000-0155-6212 R1
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06-12-2013 4/9 MFN 12-128 R1 - Description of Evaluation Page 1 of 3 Summary GE Hitachi Nuclear Energy (GEH) investigated the adequacy of a Design Change in AM 4.16-350-2C and AM 4.16-350-2H Magne-Blast Circuit Breakers as a result of a breaker failure at a PWR Licensee.
GEH has completed all testing and evaluations and has determined that the condition previously described in MFN 12-128 RO is not a reportable condition under 10 CFR Part
- 21.
This information will be sent to all GE BWR/2-6 plants and all PWRs that were previously notified under the 10CFR21.21 (a)(2) communication.
Discussion A customer reported a Functional Failure of a Safety Related Breaker (Model AM-4.16-350-2H) which involved the booster piston impacting the bottom of the booster cylinder. GEH performed a causal evaluation for the licensee and the root cause was indeterminate. It was noted that there were multiple contributing maintenance factors associated with the breaker failure.
The report concluded that no single issue could have caused the impact and subsequent failure.
GEH developed the design of this model breaker in the late 1960s which was sold to Licensees as Safety Related, prior to the development of IEEE-323. However, that did not exempt GEH from developing test reports according to existing industry standards.
One industry standard established at that time was ANSI C37.06, which required successful demonstration of 10,000 cycles under no load. This standard is used today as part of the IEEE-323 Qualification.
GEH constructed and tested a specialized breaker for a licensee in 1971 to improve response time.
Heavier springs were installed on a base model AM 4.16-350-2H.
Modification testing revealed that the heavier springs provided extra momentum to the mechanism, causing the puffer piston and booster cylinder bottom to contact.
Two additional design changes were performed to eliminate the contact. First, the opening and closing spring rates were decreased. Second, another design change lowered the booster cylinder in relation to the booster piston, providing additional impact margin (gap). Follow-up testing revealed no impact; however, the contribution of either design change is unknown.
The heavier springs were utilized for the special breaker and the product line returned to the standard springs for subsequent breaker manufacture. The modification to lower the booster cylinder was incorporated into the base product design on October 6, 1971 (with standard springs) and has carried forth until the present.
This modification increased the clearance between the piston end stroke and booster cylinder bottom.
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06-12-2013 5/9 MFN 12-128 R1 - Description of Evaluation Page 2 of 3 GEH attempted to locate documentation of the No Load Cycle Test required per ANSI C37.06 for both the pre-modification and post modification designs relating to the booster cylinder change.
GEH located documentation that demonstrates the post modification models successfully passed the ANSI C37.06/C37.09 testing in 1977, as part of the IEEE-323 Qualification. GEH was unable to locate ANSI C37.06 No Load Cycle Testing for the design prior to the 1971 modification (with standard springs).
GEH is not aware of any further design changes which would have invalidated the Qualification of the breakers prior to the IEEE-323 Qualification in 1977.
Resolution GEH determined the best method to validate the Pre-1971 design qualification was to develop a:
" Test Specification to:
o Establish the Design Basis of this portion of the breaker.
o Determine the bounding breaker types, parts and settings.
o Establish the extent of condition.
o Establishing test parameters.
(using ANSI C37.06/C37.09 as design inputs).
" Test Procedure. (Qualification Test). The purpose of the test was to:
o Validate that the previous qualification performed for the post-1971 design remains unaffected.
o Ensure qualification to IEEE-323, using ANSI C37.06/C37.09 as design inputs, is maintained.
" Analysis of Test results in regards to the impact to qualification of the Pre-1971 design.
Under the guidance of IEEE-323, "IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations," GEH has completed the testing and analysis of the test results to determine the impact to the booster cylinder change as it relates to the overall breaker qualification.
Three booster cylinders were manufactured to the Pre-1971 design and installed in a fully qualified breaker, with the bounding breaker configuration and settings. The breaker was cycled to determine the initial gap margin.
The measured gap was 5/16", which demonstrates no metal to metal contact. The gap was then trended after each 1000 cycles for a total of 5000 cycles.
Maintenance was performed per the GEH Vendor Recommendations.
No additional maintenance was performed beyond the GEH Maintenance Manual during the test. After 5000 cycles of trending, the test revealed that the gap remained consistent and margin did not degrade. The analysis performed for the test results determined:
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06-12-2013 6/9 MFN 12-128 R1 - Description of Evaluation Page 3 of 3
" The extent of condition (booster cylinder impact) was limited to breaker models AM 4.16-350-2HI2C.
This is a low wear, low fatigue portion of the overall breaker.
Sufficient Margin exists for the pre-October 1971 design.
The margin does not decrease with time (breaker cycles) when recommended GEH maintenance practices are utilized. GEH did not perform any maintenance, above that which is identified in GEH publications, in such a way that would compromise the results.
" The test and analysis confirmed the original 10,000 no load breaker test is still valid.
" There is no impact to the original ANSI C37.06/ANSI C37.09 breaker testing, which was later adopted by IEEE-323.
==
Conclusion:==
GEH has completed Testing and Analysis to validate the configuration of a Pre-1971 breaker design, which had not been previously analyzed or tested to ANSI-037.06 and ANSI-C37.09 (later invoked by IEEE-323). There is no impact to the original GEH ANSI C37.06/ANSI C37.09 breaker qualification, which was later adopted by IEEE-323. Since the test and analysis demonstrates the original breaker qualification test remains unaffected, there is no defect as defined by 10CFR Part 21, for any licensee who may have a pre-1971 manufactured breaker installed.
Technical Specification Safety Limits are unaffected as there is no means of a credible failure mechanism.
ABWR and ESBWR Design Certification Documentation Applicability The issue described above has been reviewed for applicability to documentation associated with 10 CFR 52, and determined to have no effect on the technical information contained in either the ABWR certified design or the ESBWR design in certification. This is true because the Technical Specifications submitted with the Design Certification Documentation do not include specific details associated with these components.
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06-12-2013 7/9 Page 1 of 3 US Plants Previously Notified US BWR Plants and Associated Facilities x
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x Constellation Energy Detroit Edison Co.
Energy Northwest Entergy Entergy Entergy Entergy Entergy Exelon Exelon Exelon Exelon Exelon Exelon Exelon FirstEnergy Nuclear Operating Co.
Florida Power & Light Nebraska Public Power District PPL Susquehanna LLC Progress Energy PSEG Nuclear, LLC Southern Nuclear Operating Co.
Tennessee Valley Authority Xcel Energy Plant Nine Mile Point 1-2 Fermi 2 Columbia Grand Gulf River Bend FitzPatrick Pilgrim Vermont Yankee Clinton Dresden 2-3 LaSalle 1-2 Limerick 1-2 Oyster Creek Peach Bottom 2-3 Quad Cities 1-2 Perry I Duane Arnold Cooper Susquehanna 1-2 Brunswick 1-2 Hope Creek Hatch I - 2 Browns Ferry 1-3 Monticello
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06-12-2013 8/9 Page 2 of 3 US Plants Previously Notified US P WR Plants and Associated Facilities XX X
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utility AmerenUE Arizona Public Service Constellation Energy Constellation Energy Entergy Entergy Entergy Dominion Dominion Dominion Dominion Dominion Dominion Duke Energy Corporation Duke Energy Corporation Duke Energy Corporation Exelon Exelon Exelon FirstEnergy Nuclear Operations Co.
FirstEnergy Nuclear Operating Co.
Florida Power & Light Florida Power & Light Florida Power & Light Florida Power & Light Indiana Michigan Power Corp Northern States Power Omaha Public Power District Pacific Gas & Electric Co.
Progress Energy Progress Energy Progress Energy PSEG Nuclear LLC PSEG Nuclear LLC South Carolina Electric & Gas Co.
South Texas Project Nuclear Operating Co.
Southern California Edison Co.
Southern Nuclear Operating Co.
Southern Nuclear Operating Co.
Tennessee Valley Authority Plant Callaway Palo Verde 1-3 Calvert Cliffs 1-2 Ginna Arkansas Nuclear One 1-2 Indian Point 2-3 Palisades Kewaunee Millstone 2 Millstone 3 North Anna 1-2 Surry 1-2 Waterford 3 Catawba 1-2 Oconee 1-3 McGuire 1-2 Braidwood 1-2 Byron 1-2 Three Mile Island I Beaver Valley 1-2 Davis-Besse Seabrook St. Lucie 1-2 Turkey Point 3-4 Point Beach 1-2 D C Cook 1-2 Prairie Island 1-2 Fort Calhoun Diablo Canyon 1-2 Crystal River 3 Robinson Shearon Harris Salem I Salem 2 Summer South Texas Project 1-2 San Onofre 2-3 Parley 1-2 Vogtle 1-2 Sequoyalh 1-2
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06-12-2013 9/9 Page 3 of 3 x
x x-Tennessee Valley Authority TXU Electric Generation Co.
Wolf Creek Nuclear Operating Corp.
Watts Bar I Comanche Peak 1-2 Wolf Creek