ML20195K282
| ML20195K282 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/17/1999 |
| From: | Colburn T NRC (Affiliation Not Assigned) |
| To: | Langenbach J GENERAL PUBLIC UTILITIES CORP. |
| References | |
| TAC-M99388, NUDOCS 9906220046 | |
| Download: ML20195K282 (8) | |
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.g NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 2055H001
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June 17,1999 t
Mr.' James W. Langenbach, Vice President and Director, TMl.
GPU Nuclear, Inc.
P.O. Box 480 Middletown, PA 17057
SUBJECT:
, THREE MILE ISLAND, UNIT.NO. t - ONCE THROUGH STEAM GENEPATOR (OTSG) KINETIC EXPANSION REGION INSPECTION ACCEPTANCE CRITERIA (TAC NO. M99388)
Dear Mr. Langenbach:
During the 11R refueling outage at Three Mile Island, Unit No.1 (TMI-1), NRC inspectors identified that the steam generator tube inspection practices of GPU Nuclear, Inc., (GPUN) were atypical from general industry approaches.. Specifically, GPUN was applying a voltage-based repair criteria to disposition indications of intergranular attack (IGA) initiating from the inside diameter of the tubing. In add' tion, GPUN did not appear to have a program in place to
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.i ensure the continued structural or leakage integrity of upper tubesheet kinetic expansion joints j
. containing IGA indications. The NRC and representatives from GPUN held meetings on
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November 26,1996, and July 25,1997, to discuss GPUN inspection practices with regard to
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criteria established to address the IGA degradation. During the meeting held on July 25,1997, glPj GPUN representatives agreed to submit the technical basis for inspection ar d rupair criteria that were to be applied to indications in the upper tubesheet kinetic expansion joints in the 12R
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outage _for TMI-1. The staff received the GPUN structural assessment of degraded tubesheet 1
expansions ~on August 8,1997, in accordance with the staff's request in the July 25 meeting.
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The leakage integrity assessment was obtained subsequent to the TMI-1 refueling outage on j
- November 26,1997.
1 Subsequent to the meetings held with GPUN in 1996 and 1997, the staff conducted a review of the repair criteria applied in the 12R outage including an in-depth assessment of the regulatory requirements applicable to the degradation in the upper tubesheet region. As explained below, the sta*f has determined that the repair criteria implemented by GPUN require regulatory '
approval prior to implementation.
According to the GPUN August 8, and November 26,1997, submittals, the steam generator
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tube repair criteria included in the surveillance requirements of the TMI-1TSs do not apply to the upper and lower tubesheet regions. The basis for this conclusion stems from the L requirements in TS 4.19.4.a.8 which indicate that the bounds for inspections are from the
, secondary face o - ach tubesheet. Therefore, indications of degradation identified outside of
' this region do no', pear to be subject to the depth-based repair criteria included in TS - 4.19.4.a.6, that is, plug or repair degradation with depths exceeding 40 parcent of the nominal.
tube wall thickness.
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'9906220046 990617 PDR ADOCK 05000289 P
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.J.Langenbach The NRC agrees with GPUN that the TMI-1 TSs do not appear to apply to degradation within the tubesheet regions. However, in the absence of inspection and repair criteria in the surveillance requirements, the nicensee is required to default to criteria established for the reactor coolant pressure boundary (i.e., American Society of Mechanical Engineers Boser and s
Pressure Vessel Code (ASME Code) in accordance with 10 CFR Part 50.55a. lWB-3521.2,
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" Allowable Flaws for Straight-Tube Steam Genewtors," within Section XI of the ASME Code,
' currently contains no established acceptance criteria applicable to the tubes in the TMI-1 or other once-through steam generators. In the absence of flaw acceptance criteria, GPUN hould develop and submit the basis for the acceptance criteria r.pplied at TMI-1 in accordance s
with lWB-3630, " Acceptance Criteria For Steam Generator Tubing." Alternatively, the licensee could employ the depth-based repair criteria cunsntly included in the TMl-1 TSs. A depth-1
- based approach appears to have been used by GPUN prior to the chemistry transient the.t resulted in the development of defects in the upper tubesheet region in the SR outage. This 4
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statement is supported by references included in the documentation used to support the original repair of the tubes after the IGA degradation was detected in November of 1981. The staff notes that other pressurized-water reactor utilities with similar (to TMl-1) inspection boundary i
requirements in their plant TSs have adopted the repair limits and associated requirements j
applicable to tube' areas within the inspection boundary defined in the TSs to any areas which
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are outside the TSs defined boundary, in this regard, the staff concludes that the development of criteria that differs from the requirements of TS 4.19.4.a.8 is inconsistent with industry practice. The staff recognizes that GPUN is permitted to adopt attemate criteria provided it has been approved by the NRC.
j The information supplied by your August 8 nrd November 26,1997, letters constitutes the technical basis for the repair criteria applied to indications in the upper tubesheet region in the
- _ most recent refueling outage. The taaff has reviewed this information and identified several issues that require further clarification in order for the staff to draw a conclusion regarding the acceptability of the criteria. developed by the licensee. The specific issues in need of clarification are included in the Enclosure to this letter. The NRC requests that the liccasee
' provide this information within 30 days of the date of receipt of this letter. As discussed in the teleconw.c on June 10,1999, with Mr. Knight of your staff and others, revised criteria have been developed for use during the upcoming TMl-1 refueling outage (13R). The staff also requests that you submit for review, a description of, and the technical basis for the revised inspection criteria for the 13R outage as soon as practical following development and
- completion of your intemal review (we understand this to be early August 1999). The NRC a
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June 17,1999 1
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i staff will review the acceptance criteria in accordance with IWB-3630 and provide you with a safety evaluation documenting our conclusions on the repair criteria shortly thereafter. Should a situation occur that prevents you from meeting the target date, or if you would like to further l
d.:scuss the information requested in this letter, please contact me,301-415-1402.
Sincerely, Original signed by:
Timothy G. Colburn, Senior Project Manager, Section 1 Project Directorate l Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-289
Enclosure:
Request for Additional Information cc w/ enclosure: See next page DISTRIBUTION:
SeekstmeEP; TClark SBlack TColburn PUBLIC OGC PEselgroth, RI PDI R/F ACRS PRush WBateman l
l DOCUMENT NAME: G:\\PDI-2\\TMI-1\\99388RAl.WPD
. To receive a copy of this document, Indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with e ttachment/ enclosure "N" = No copy 0FFICE PM:PDI-1 % lE pfDI-1M l
SC:EMCB l
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l NAME TColburn/mpw Elirk ~ ~ GM ESu111 van [ S SBajwa # M d DATE 06//f/99 06/15 /99 06/[6/99 0
06//7 /99 06/
/99 Official Record Copy i
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June 17,1999 i
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J.Langenbach
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staff will review the acceptance criteria in accordance with IWB-3630 and provide you with a safety evaluation documenting our conclusions on the repair criteria shortly thereafter. Should a situation occur that prevents you from meeting the target date, or if you would like to further discuss the information requested in this letter, please contact me, 301-415-1402.
Sincerely, Original Signed by:
Timothy G. Colburn, Senior Project Manager, Section 1 Project Directorate i Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-289
Enclosure:
Request for Additional Information cc w/ enclosure: See next page DISTRIBUTION:
Docket File TClark SBlack TColburn PUBLIC OGC PEselgroth, RI PDI R/F ACRS PRush WBateman DOCUMENT NAME: G:\\PDI-2\\TMI-1\\99388RAl.WPD To receive a copy of this document, indicate in the box: "C" = Copy without attachment /enclosurc "E" = Copy with rttachment/ enclosure "N" = No coay 0FFICE PM:PDI-1 % lE M;,PDIld l
SC:EMCB l
SC:PDI.1 l
l NAME TColburn/mpw
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ESullivan f M SBajwa # 362.
DATE 06//f/99 06/l5/99 06/l.6/99 0
06//7/99 06/
/99 Official Record Cop.y 4
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E7 J.Langenbach
-3, staff will review the acceptance criteria in accordance with IWB-3630 and provide you with a safety evaluation documenting our conclusions on the repair criteria shortly thereafter. Should a situation occur that prevents you from meeting the target date, or if you would like to further discuss the information requested in this letter, please contact me, 301-415-1402.
Sincerely, 7~
W Timothy G. Co burn, Senior Project Manager, Section 1 Project Directorate l Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-289
Enclosure:
Request for Additionalinformation cc w/ encl: See next page w.
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s.p Three Mile Island Nuclear Station,' Unit No.1 cc:
- Michael Ross.
Robert B. Borsum Director, O&M, TMI
- B&W Nuclear Technologies
. GPU Nuclear, Inc.
Suite 525 P.O. Box 480 1700 Rockville Pike Middletown, PA 17057 Rockville, MD 20852 John C. Fomicola -
William Dornsife, Acting Director Director, Planning and Bureau of Radiation Protection
' Regulatory Affairs Pennsylvania Department of GPU Nuclear, Inc.
Environmental Resources 100 Interpace Parkway P.O. Box 2063
~ Parsippany, NJ 07054 -
Harrisburg, PA 17120 Jack S. Wetmore Dr.' Judith Johnsrud Manager, TMl Regulatory Affairs National Energy Committee GPU Nuclear, Inc.
Sierra Club P.O. Box 480 433 Orlando Avenue Middletown, PA 17057 State College, PA 16803 Ernest L. Blake, Jr., Esquire Peter W. Eselgroth, Region l Shaw, Pittman, Potts & Trowbridge U.S. Nuclear Regu!atory Commission 2300 N Street, NW.
Washington, DC 20037_
475 Allendale Road
' King of Prussia, PA 19406 Chairman J
Board of County Commissioners
' of Dauphin Ccunty Dauphin County Courthouse Harrisburg, PA 17120.
Chairman Board of Supervisors _
l of LondonderryTownship R.D. #1, Geyers Church Road Middletown, PA 17057 -
Wayne L Schmidt
, Senior Resident inspector (TMI-1) 4
- U.S. Nucinr F.egulatory Commission P.O. Box 219 Middletown; PA 17057
~
l Regional Adminietrator '
i Region 1 -
. U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 i
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REQUEST FOR ADDITIONAL INFORMATION OTSG KINETIC EXPANSION INSPECTION ACCEPTANCE CRITERIA 1.
The axial, once-through steam generator (OTSG) tube loads used in the structural assessment were considered without including a factor of safety. The staff has identified that the use of nominal, thermally-induced loads may not yield conservative results in situations involving large thermal displacements imposed on flawed steam generator tubes. In addition, the staff has been informed that the limiting OTSG tube loads applicable to TMI-1 are currently being evaluated by Framatome Technologies, incorporated (FTI). Preliminary results from FTI's analyses indicate that the loads considered in MPR-1820, Revision 0 (proprietary), are greater than the peak axial loads applicable to OTSG tubing.' However, it is unclear if the loads considered by the licensee bound those that wik result from FTI's study when considering additional factors of safety. Discuss whether the structural repair criteria developed for the 12R outage remain valid considering the revised tube load analyses conducted by FTl and the application of factors of safety to these loads. Provide the technical bases for any safety factors considered in this assessment.
2.
Section 3.6.1 of MPR-1820, Revision 0, states that the inspection acceptance criteria for tube.* at any location can be determined by interpolating between the results for the center, mid-radius, and peripherallocations. The staff notes that this approach may result in nonconservative flaw sizes for 22-inch expansions for tubes within a relative radii between 0.707 and 1. Provide additional Information to explain how the repair criteria remain conservative in this situation.
3.
The structural acceptance criteria included in Table 3-5 was determined by analyses that assumed a primary-to-secondary differential pressure of 2500 psi acting simultaneously with the peak axial tube loads. This assumption appears to be invalid considering the sequence of the thermal and pressure loads of Topical Report #116, Revision 0, submitted by the licensee on November 26,1997. Including a pressure load cf 2500 psi in the analyses will result in higher than expected tube to tubesheet contact pressures due to pressure effects at the moment when axial tube loads are at their greatest. As a result, the defect length criteria may be nonconservative. D;scuss the technical basis for assuming an applied differential pressure of 2500 psi in conjunction with the largest axial tube loads in the development of the acceptance criteria.
~ 4.
- it was stated in the submittal dated November 26,1997, that the leakage assessment would only consider potential leakage from flaws with a measured through-wall depth greater than 67.4 percent. The staffis unaware of any qualified depth sizing technique..
. for the degradation of interest in the TMI-1 OTSGs. Provide the technical bases for assuming that flaws with a measured through-wall depth of less than 67.4 percent will
' not leak during a main steam line break accident.
t ENCLOStJRE
i 2-5.
The technical bases for the steam generator tube repair criteria that were submitted to the staff on August 8 and November 26,1997, indicated that these criteria would be used in the 12R outage. However, the submittals did not indicate whether these criteria 3
would be used in subsequent steam generator tube inspections at TMI-1. Discuss whether the inspection and repair methodology developed to address indications in the kinetic expansions in the 12R outage will be used in subsequent inspections. If tne licensee intends to develop new criteria or revise that discussed in previously submitted documents, the staff requests that the licensee submit a description of its technical bases for the new criteria similar to that provided for the 12R inspections for NRC review and approvalin accordance with IWB-3630 in Section XI of the ASME Code.
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