ML20141J398
| ML20141J398 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 08/12/1997 |
| From: | Langenbach J GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20141J403 | List: |
| References | |
| 6710-97-2357, NUDOCS 9708200040 | |
| Download: ML20141J398 (11) | |
Text
- ___ __________-_-____--___----_ ----- - - - - - -
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4 l GPU Nuclear,Inc.
( Route 441 south NUCLEAR P**'0c'******
Middletown, PA 17057 0480 Tel717 944-7621 6710-97-2357 August 12, 1997 U. S. Nuclear Regulatory Commission
! Attention: Document Control Desk-Washington, DC -20555
Dear Sir:
l I
Subject:
Three Mile Island Nuclear Station, Unit 1, (TMI-1)
Operating License No. DPR-50 Docket No. 50-289
)
Technical Specification Change Request (TSCR) No. 268 l
In accordance with 10 CFR 50A(b)(1), enclosed is TSCR No. 268. Also enclosed is the Certificate of Service for this request certifying service to the chief executives of the township and county in which the facility is located, as well as the designated ollicial of the Commonwealth of-Pennsylvania, Bureau of Radiation Protection.
The purpose of this TSCR is to request changes to the Surveillance Specification for Once Through Steam Generator (OTSG) inservice inspection for TMI-l Cycle 12 Refueling (12R)
- examinations applicable to TMI-l' Cycle 12 operation.
GPU Nuclear, Inc. requests that the amendment authorizing this change become effective prior to startup following the 12R Outage which is scheduled to begin on September 5,1997. Since the additional requirements to be implemented by this proposed amendment request are more restrictive than the current technical specification requirements and GPU Nuclear is committed to compliance with the more restrictive requirements including in-situ pressure testing during the 12R Outage, we do not believe that the requested amendment should be a pre-requisite to startup following the 12R Outage. However, if the amendment proposed by this request is to be required 9708200040 970t312" p '
PDR ADOCK 05000299 P
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6Il0-97-2357; Page 2 of 2:-
- prior to Cycle 12 operation, GPU Nuclear requests that an amendment be issued on or before October 3,1997J Sincerely, .
%tiV $4 Y ames W. Langen h Vice President an irector, TMI Attachments L M.RK -
l - cc: Administrator NRC Region 1-L TMI Senior NRC Resident inspector l TMI-l Senior NRC Project Manager
'
- METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND
. PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION, UNIT 1 Operating License No DPR-50 Docket No. 50-289 Technical Specification Change Request No. 268 COMMONWEALTH OF PENNSYLVANIA )
) SS COUNTY OF DAUPHIN )
This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station, l Unit 1. As a part of this request, proposed replacement pages for Appendix A are also included. '
All statements contained in this submittal have been reviewed, and all such statements made and matters set forth therein are true and correct to the best of my knowledge.
GPU NUCLEAR, INC.
BY:
tLLb' l $1'A&{
ector, TMI
[ ice Presidenthd Sworn,pnd Subscri to before me this/,4 day of / M ,1997.
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/' "" ' * ""6c'
' Mn'ral Seal ' ' N' Suzanne C. MAlcah, Notary PuMc My o miss?on hs av , Ib9 Member, Pennsylvania Assouatwn of Notata
. _, UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION INTHE MATTER OF DOCKET NO. 50-289
, - GPU NUCLEAR INC. LICENSE NO. DPR-50 CERTIFICATE OF SERVICE '
This is to certify that a copy of Technical Specification Change Request No. 268 to Appendix A L of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with executives of Londonderry Township, Dauphin County, Pennsylvania; Dauphin County, Pennsylvania; and the Pennsylvania Department of Environmental Resources, Bureau ofRadiation Protection, by deposit in the Unitet. States mail, addressed as follows:
Mr. Darryl LeHew, Chairman Ms. Sally Klein, Chairman
. Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Courthouse Middletown, PA 17057 Harrisburg, PA 17120 Director, Bureau of Radiation Protection PA Dept. of Environmental Resources Rachael Carson State Office Building -
P.O. Box 8469 Harrisburg, PA 17105-8469 Att: Mr. Stan Maingi GPU NUCLEAR INC, 1 S S I Eesident and _ tor, TMi DATE: //wd /d -
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710-97-2357 Attachment Page 1 of 6 1.
TECHNICAL SPECIFICATION CHANGE REQUEST (TSCR) NO. 268:
GPU Nuclear (GPUN) Inc. requests changes to the TMI-l Technical Specifications, on the replacement pages designated below, as follows:
Please insert the revised replacement pages 4-79 through 4-83 which are enclosed with this submittal.
II. REASON FOR CHANGE:
The current Technical Specification 4.19, "OTSG Tube Inservice Inspection," reflects a state of the art eddy current technology and understanding of that technology from the mid 1980's following the sulfur intrusion event which caused damage to the TMI-l Once Through Steam Generators (OTSGs) in 1981, the kinetic expansion repairs which were performed in 1982, and subsequent testing. Testing following the kinetic expansion i repairs was evaluated in NUREG-1019, " Safety Evaluation Report related to Steam Generator Tube Repair and Return to Operation of Three Mile Island Nuclear Station, Unit No.1" and NUREG-1019, Supplement No.1, which was part of the NRC's Safety Evaluation Report (SER) for TMI-l License Amendment No.103, dated December 21,1984. Although GPU Nuclear has continued to incorporate improvements in eddy current technology to enhance the detection of tube degradation, the current technical specifications require updating based on current knowledge of the degradation that has been detected at other plants and the limitations of the eddy current techniques for through wall sizing ofInter-Granular Attack (IGA) degradation.
GPU Nuclear met with the NRC on July 15,1997 to discuss NRC concerns with the below voltage criteria (BVC) eddy current test (ECT) indications regarding compliance with the current TMI-l Technical Specification (TS). During the meeting GPU Nuclear described the plans for the performance of eddy current examinations during the 12R Outage and the acceptance criteria to be used, which go beyond the current technical specifications and address the NRC's concerns. Following the presentation the NRC requested that the additional terms and criteria be included in a Technical Specification Change Request (TSCR) for tube examinations that are to be performed during the next outage,12R, which it. scheduled to begin on September 5,1997. The NRC stated that the changes that were discussed would apply for one cycle of operation only, Cycle 12, and that long term permanent changes would later be needed to address the requirements of a future generic letter (GL) and regulatory guide (R. G.), which are presently in preparation. The NRC expects to publish the GL and R. G. in time for incorporation of additional changes into the TMI-l Technical Specifications prior to the TMI-l Cycle 13
, Refueling (13R) Outage, which is scheduled to begin in September 1999.
- 6710-97-2357 Attachment Page 2 of 6 III. SAFETY EVALUATION JUSTIFYING CHANGE These proposed changes impose axial and circumferential extent sizing limitations in addition to existing Technical Specification requirements for Inside Diameter (ID) initiated degradation where bobbin coil ECT signal amplitudes do not permit reliable ]
through wall sizing. By applying these additional axial and circumferential sizing limits to all ID IGA (regardless of through wall depth or amplitude) additional assurance is provided in the technical specifications that structurally significant degradation will be removed from senice or repaired and will not result in tube failure or excessive leakage I during the limiting postulated Main Steam Line Break (MSLB) accident. Additional criteria have been provided to conservatively identify degraded tubes for C-1, C-2, or C-3 inspection results categorization.
The probability of OTSG tube leakage is not adversely affected by the proposed dispositioning strategy for inside diameter initiated IGA. Operating history indicates essentially no primary-to-secondary leakage through insenice OTSG tubes at TMI-1.
Growth rate studies indicate that this damage mechanism has been and is expected to continue to be inactive. TMI-l's operating license currently restricts allowable primary-to-secondary leakage to < 0.1 gpm above baseline (currently zero) and is consistent with the guidance in Generic Letter (GL) 95-05, " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking." This dispositioning strategy ensures that inside-diameter initiated volumetric eddy current indications exceeding the length criteria are repaired or removed from service during the tube insenice inspection. These criteria provide additional assurance of OTSG tube integrity.
The following describes the changes as they appear on the revised technical specification pages. Note that because of the text that has been added, the expanded text has caused some of the revised sections to be moved to the next page. The following refers to the page numbers associated with location of text on the revised pages:
Page 4-79:
Added to NOTE 1 in Section 4.19.2, " Steam Generator Tube Sample Selection and Inspection," is the stipulation that previously degraded tubes whose degradation has not been spanned by a sleeve must exhibit significant increase in the applicable degradation size measurement to be included in the percentage calculations that apply to the inspection results categories C-1, C-2, or C-3). This is an administrative change to update the technical specifications, subsequent to the installation of sleeves during TMl-1 Outages 9R and 10R, and acknowledge that degradation spanned by a sleeve is already repaired.
In addition to the criterion that previously degraded tubes with greater than 10%
increase in through wall depth size measurement be used in categorizing the inspection results, this change includes an additional criterion of bobbin coil amplitude
- 710-97-2357 Attachment Page 3 of 6 greater than 0.6 volt for the ID IGA degradation that cannot be depth sized but are continuing to degrade beyond ECT repeatability variations for use in the categorization of C-1, C-2, or C-3 inspection results. The 0.6 volt amplitude limit is based on a two sigma standard deviation of the average of the ID IGA statistical growth evaluations for the past 7 outages (See Table 111-6 in the GPU Nuclear Cycle 11 Refueling (11R) Outage OTSG Tube Inspection Report, which was submitted to the NRC on August 1,1996).
Editorial changes to improve the consistency of format include moving the indent for the text following the 4.19.3 section heading and adding a period at the end of subsection 4.19.3.b.
Page 4-80:
Section 4.19.4, " Acceptance Criteria," subsection a.1, defines an Imperfection. In place of the statement that eddy current testing indications below the 10% degraded tube criteria of the current technical specifications may be considered imperfections, the words "20% of the nominal tube wall thickness, if detectable," are being revised.
The revised words remove the reference to through wall extent and refer to the new
" degraded tube criteria specified in 4.19.4.a.3 below." As with the current technical specifications, indications that are less than the degraded tube criteria are thereby considered imperfections.
Section 4.19.4, " Acceptance Criteria," subsection a.3, defines the criteria for a Degraded Tube. In addition to the present degraded tube criteria (imperfections 2 20% of the nominal tube wall thickness caused by degradation), new acceptance criteria are added. The revised text states that these criteria are applicable for one cycle only. In accordance with the proposed changes for the Cycle 12R Outage examinations, tubes containing "an inside diameter (l.D.) IGA indication with a bobbin coil indication 2 0.5 volt or 2 0.13 inches axial extent or 2 0.28 inches !
circumferential extent" will be considered degraded. This adds criteria to consider tubes with ID IGA that cannot be through wall depth sized so that they may be included in degraded tube totals for categorization as C-1, C-2, or C-3 inspection results.
Page 4-81:
Section 4.19.4, " Acceptance Criteria," subsection a.6, defines the Espair Limit criteria. The proposed changes incorporate a " limit for inside diameter IGA indications" of"0.25 inches axial extent or 0.57 inches circumferential extent," in addition to the current 40% nominal through wall limit. An assumed 100% indication 1 with axial extent of 0.25 inches or circumferential extent of 0.57 inches has been evaluated to meet the R. G.1.121 acceptance criteria for margin to failure for the postulated MSLB applied differential pressure and axial tube loads. These revised criteria provide a high confidence that unacceptable flaws that do not provide the
.+
6710-97-2357 Attachment Page 4 of 6 required structural integrity to withstand a postulated MSLB accident will be remove.d from service. Based upon the results of growth rate evaluations which indicate that the inside diameter IGA mechanism is inactive, the probability of tube leakage from an ID IGA flaw due to an accident at the end of the operating cycle is essentially unchanged from the probability at the beginning of the cycle.
Section A 'M Acaptance Criteria," incorporates a new subsection a.9, for "Inside Diameter InJmVranular Attack (IGA) Indication," defmed as "a bobbin coil indication initiating on the inside diameter surface and confirmed by diagnostic ECT to have a volumetric morphology characteristic ofIGA." This is a new criterion which ensures that the acceptance criteria for degraded tubes will only be applied to patch-like IGA indications and not to those that may be indicative of a crack.
Section 4.19.5, "RepmLs," is being revised to include an assessment ofID IGA l
degradation growth and the results of any in-situ pressure testing performed during the outage in addition to providing the number of tubes repaired or removed from service. Rather than providing this in a report within 15 days, this change requires that the NRC be notified of the above prior to exceeding a Reactor Coolant (RCS) temperature of 250 'F. j Egge 4-82:
Section 4.19.5, " Reports," subsection 4.19.5.b, includes a change to shorten the required schedule to 90 days for providing the complete results of steam generator tube inspections and repairs. This schedule ensures that the NRC is provided with a report of the complete results much sooner than the 12 month schedule which is currently specified. The words "and repairs"is added for clarification to define the start of the time period allowed for preparation of this report. In addition to that information that is currently required to be included in the report of the complete results of these inspections, a requirement is being added. The additional information includes the " Location, bobbin coil amplitude, and axial and circumferential extent (if determined) for each inside diameter IGA indication."
Page 4-83:
This page incorporates changes to the Bases which relate to the requested changes in Section 4.19, "OTSG Tube Inservice Inspection specifications."
6710-97-2357 Attachment Page 5 of 6 IV. NO SIGNIFICANT HAZARDS CONSIDERATION:
GPU Nuclear has determined that this TSCR poses ao significant hazards consideration as defined by 10 CFR 50.92.
A. These proposed changes do not represent a significant increase in the probability of occurrence or consequences of an accident previously evaluated. The only accidents
- previously evaluated that could be sil mificantly affected by changes to the OTSG tube insenice inspection requirements are the steam generator tube rupture (STGR) and !
the main steam line break (MSLB) accidents.
The proposed flaw disposition strategy based on measurable eddy current parameters of axial and circumferential extent for Inside Diameter (ID) Initiated Inter-Granular Attack (IGA) will provide high confidence that unacceptable flaws that do not have the required structural integrity to withstand the MSLB are removed from senice.
The proposed axial and circumferential length limits for eddy current inside diameter degradation indications meet the RG 1.121 acceptance criteria for margin to failure for MSLB applied differential pressure and axial tube loads. The capability for detection of flaws is unaffected and the identification of tubes which should be repaired or removed from senice is maintained or improved. The operation of the OTSG or related structures, systems, or components is otherwise unaffected.
Therefore, neither the probability nor consequences of a SGTR is significantly increased either during normal operation or due to the limiting loads of a MSLB accident.
Neither the editorial changes in format, punctuation, or grammar nor the administrative changes or changes in reporting requirements, as described above, could significantly affect the probability of occurrence or consequences of any accident previously evaluated.
B. These proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated because there are no hardware changes involved nor changes to any operating practices. These changes involve only the OTSG tube inservice inspection surveillance requirements, which could only affect i
the potential for OTSG pr mary-to-secondary leakage. The proposed changes impose additional flaw length limits for ID IGA that go beyond existing requirements to assure tube structural and leakage integrity.
In addition, neither the editorial changes in format, punctuation, or grammar nor the administrative changes, as described above, could possibly create the possibility of an accident of a new or different type from any previously evaluated. These changes are included only to improve the clarity and readability of the Technical Specifications and comply with the NRC's desire to obtain the results of the inspections as soon as practical.
- 6710-97-2357 Attachment Page 6 of 6 Therefore, these changes do not create the potential for single or multiple tube ruptures or any other kind of accident different from those that have been evaluated.
C. These proposed changes do not involve a significant reduction in a margin of safety because the changes am more restrictive than the current technical specification and the margins of safety defined in R.G.1.121 are retained. The probability of detecting degradation is unchanged since the bobbin coil eddy current methods will continue to be the primary means ofinitial detection ar 1 the probability ofleakage from any indications left in service remains acceptably small. The strategy for dispositioning ID initiated IGA will continue to provide a high level of confidence that tubes exceeding the allowable limits for tube integrity are repaired or removed from service.
In addition, neither the editorial changes in format, punctuation, or grammar nor the administrative changes or changes in reporting requirements, as described above, l could significantly affect a margin of safety and are included only to improve the clarity and readability of the Technical Specifications and comply with the NRC's j desire to obtain the results from tube inspections as soon as practical, t
V. IMPLEMENTATION:
GPU Nuclear, Inc. requests that the amendment authorizing this change become effective prior to startup following the TMI-l Cycle 12 Refueling (12R) Outage which is scheduled to begin on September 5,1997. Since the additional requirements to be implemented by this proposed amendment request are more restrictive than the current technical specification requirements and GPU Nuclear is committed to compliance with the more restrictive requirements including in-situ pressure testing during the 12R Outage, we do not believe that the requested amendment should be a pre-requisite to startup following the 12R Outage. However, if the amendment proposed by this request is to be required prior to Cycle 12 operation, GPU Nuclear requests that an amendment be issued on or before October 3,1997, 1
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Revised Technical Specification Pages
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