ML20085H269
| ML20085H269 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/16/1983 |
| From: | Wagner E GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20083L846 | List: |
| References | |
| NUDOCS 8309070458 | |
| Download: ML20085H269 (10) | |
Text
r-REPORT OF THE THIRD PARTY REVIEW OF THREE MILE ISLAND, UNIT 1, STEAM GENERATOR REPAIR SUPPLEENT 1 To R. F. Wilson - Vice President, Technical Functions GPU Nuclear Prepared by THIRD PARTY REVIEW GROUP:
Stephen D. Bio'wn Stanley A. Holland Arturs Kalnins William H. Layman David J. Morgan Richard W. Weeks Edwin J. Wagner - Chairman Submitted for the Review Grou by:
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3309070458 830825 PDR ADOCK 05000299 P
i PURPOSE:
This report supplements the report of the Third Party Review of the TMI-l steam generator repair dated February 18, 1983. The February 18 report contained findings, conenents and reconsnandations to which GPU Nuclear responded. The GPU Nuclear written responses were further described and evaluated in a Review Group meeting of April 12 and 13, 1983. This supplementary report is intended to be the final report l
by this Review Group.
BACKGROUND:
The establishment and operation of the Third Party Review Group of the Tr;I-l steam generator repair was previously discussed in the Review Group's Report of February 18,1983. The report contained a conclusion concerning the adequacy of the steam generator repairs for safe operation of the TMI plant and findings, connents and reconnendations about the steam generator repair and return of the plant to operation.
GPU Nuclear responded to the Review Group's report in an April 7,1983 letter and submitted additional documents for the Review Group's use.
The April 7,1983 letter with its attachment 1 is included as Appendix A with this report. Also GPU Nuclear letter of April 4,1983, which distributed additional documents for the Review Group's use, is Appendix 8.
The focus of the Review Group's meeting on April 12 and 13,1983 was the GPU Nuclear responses in Attachment 1 of Appendix A and the revise"i
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Safety Evaluation Reports 008 and 010. During this meeting, GPU i.aclear made presentations and responded to the Review Group's questions. The Review Group determined the adequacy cf the GPU Nuclear responses and modified its conclusion, r
At the close of this Review Group meeting, its revised conclusion was presented to Mr. R. F. Wilson, Vice President, Technical Functions, Mr. P. R. Clark, Executive Vice President, GPU Nuclear staff and an NRC staff representative.
CONC 1.USION:
The February 18, 1983 report concluded that it would be premature to detennine that, when all existing GPU Nuclear plans are completed, the Third Party Review would conclude that the results will be positive and will ensure that plant operation would be without increased risk.
Based upon the revised documents identified in Appendix A and 8, and particularly the Safety Evaluation Reports 008 and 010, the Review Group modified its prior conclusion. The identified documents described substantial additional analyses and testing done by GPU Nuclear and its contractors on activities identified on the Review Group's prior report as' incomplete including:
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Post repair testing and hot functional operation of the l
systems.
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Completion of analyses including leak before break and the contingency of multiple tube rupture.
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Translation of analytical work such as leak before break and multiple tube rupture into useable plant guidance, procedures and training.
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A conservative approach of power. escalation after completion of repairs.
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The Review Group now concludes that, upon satisfactory congletion of the entire program as defined in the safety evaluations and as augmented by l
GPU Nuclear comments during and subsequent to the April 12 and 13 meeting, the TMI-l plant can be operated safely with tha repaired steam generators.
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COM9ENTS AND RECOMENDATIONS:
The Review Group specifically addressed each of the GPU Huclear responses of Appendix A and determined the acceptability of the responses.
In the following section each of the GPU Nuclear responses to earlier Review Group connants and reconmendations are addressed. The same format and numbering are followed as in Appendix A.
The GPU Nuclear responses are not rettated here. Refer to Appendix A.
A.
Steam Generator Inspection Reconnendation 1 - The difference between the Review Group reconnendation and the GPU Nuclear response concerned 17 tubes total in both steam generators. The 17 tubes have eddy current indications of less than 40 percent through wall and over two coils in extent.
The Review Group had recommended plugging these tubes but GPU Nuclear responded that the 17 tubes would remain unplugged to provide information on crack growth between periodic eddy current examinations.
The Review Group considers the GPU Nuclear response to be satisf:ctory.
It is noted that the indication size is substantially less than the critical crack size developed in Safety Evaluation Report 008 and thus would not present a safety risk.
Reconnendation 2 - The Review Group previously reconnended the plugging of all tubes with eddy current indications at or above the 15th tube support plate within three rows of the tube lane. GPU Nuclear responded that these tubes are or will M plugged except that tubes in the second or third row from the tube lane will not be plugged.
The Review Group was advised in the Aoril 12 and 13 meeting that the extent of participation of tubes in the tube lanes was evaluated based upon prior eddy current examination of TMI-1 steam generators. Only tubes in the first row from the tube lane have
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shown indications associated with tube lane vibration. Based upon this specific infonnation about vibration perfonnance of these steam generators, the Review Group considered this response satisfactory.
Reconnendation 3 - The GPU Nuclear response is satisfactory.
Reconnendation 4 - The GPU Nuclear response is satisfactory.
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B.
Cause of Tube Cracking Reconnendation 1 - The Review Group had reconnended earlier that GPU Nuclear implemer.t corrective measures or verify their current programs for minimizing the ingress of all impurities (not just sulfur) into the reactor coolant system. The response addressed actions to protect from impurities. Although the GPU Nuclear actions are considered adequate for safety, the Review Group 1
nade the following connants concerning impurity control and related chemistry program.
Further Connent 1 - Identification of Sources - GPU Nuclear presented their plans for a chemical control program to be implemented over the next year or so. This program would identify i
l paths for inpurity input to the reactor coolant system and would impose controls to minimize input. We. commend this effort and reconnend that it be accelerated if possible.
Further Connent 2 - Resin Loss from Purification Ion Exch' angers -
We concur with GPU Nuclear's assessment that resin loss will be detected by pressure drop increases across the reactor coolant pump seal water inlet filters. These filters are of small pore size and receive a large fraction of the flow from the ion exchangers. GPU Nuclear's plans for changing out the purification ion exchanger post filters during the startup should permit identification of any physicrl problems and should reduce the potential for impurity input during the proposed peroxide cleanup.
We concur that specific sampling and analyses for resin fines is not necessary.
j Further Connent 3 - Control of Organics Input - Make-up water analyses presently specified will not detect organics. These materials can contain sulfur, chlorine, and other aggressive impurities which will be released to reactor coolant under heat and radiation. Also, carbonaceous material was found to be the major impurity near tube failure, and may have played a role in the failure which, in our ignorance, we do not understand.
For these reasons, we reconnend that specific analyses for organics be performed on the make-up water and other input streams to the reactor coolant system. GPU Nuclear indicated that the were in the process of purchasing a Total Organic Carbon (TOC)y analyzer. This purchase should be expedited and analysis for TOC should be added to the Impurity Ingress Control Program. An initial guideline of 1 PPM TOC was suggested.
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4-Further Comment 4 - Control of Recycle Streams - We understand that TMI-1 does not. plan to recycle water or boric acid recovered from the Radioactive Waste Processing Systems, although the design pennits it. We reconnend thet plant procedures be reviewed to assure that recycle is not an option until the recycle streams have been incorporated into the Impurity Ingress Control Program.
Further Connent 5 - Chemistry Specifications - With the exception below on sodium limit, the Review Group concurs with the chemistry control program planned by GPU Nuclear. Our exception on the sodium limit is not considered to be of rignificance to plant safety.
Further Connent 6 - Reactor Coolant Sodium Limit - GPU Nuclear has established a limit of 1.0 ppm for scdium in reactor coolant. The review Group suggests that this be reduced an order of magnitude to 0.1 ppm.
With the removal of the sodium thiosulfate addition system, sodium is no longer likely to be present in the reactor coolant system.
The purification ion exchangers are easily capable of maintaining this limit, even in plants which use sodium.
hydroxide to adjust the pH of reactor building spray (providing a possible source for sodium into reactor water).
Sodium of 1 ppm will noticeably affect conductivity and pH, complicating the reconciliation of these measurements with lithium and boron concentrations.
Lower sodium concentrations will be reflected in lower Na -24 activity in the coolant and waste streams.
Sodium should be used as an indicator of potential problems with the purification ion exchangers or in one of the make-up streams (demineralized water, concentrated boric acid, etc.)
to the reactor coolant system. As such, its limit should be only slightly above its nominal concentration in the reactor coolant system. The nominal or " background" sodium should be determined from plant experience.
C.
Materials Application Connent 1 - The Review Group previously cautioned GPU Nuclear to remain alert to the possibility that small cracks may have gone undetected in the reactor coolant system. Since then, additional inspections in the waste gas system and the pressurizer have located additional evidence of high sulfur concentrations and corrosive attack. Our previous note of caution continues to be valid although the Review Group does not consider further inspections to be necessary at 4
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this time.
Comnents 2 and 3 The Review Group had connented previously on the sparcity of crack propagation rata data for the steam generator tubes, and the subsequent large extrapolation of the data required for the conditions of interest.
Since then, more data were found to help substantiate GPU's analysis, although extrapolation is still required. One of the conclusions of the most recent GPU Nuclear analysis is that flow-induced vibrations say not play any role in propagating steam generator cracks.
Nevertheless, if practical, for conservatism the Review Group still suggests that the long-tenn corrosion tests, ~which are designed to anticipate problems before their occurrence in the plant, should include a simulated flow-induced vibration loading. By simulating the actual loading, material conditions,.and expected environment as closely as possible, these tests should help to warn of any problem that may have been missed in the analysis.
Connent 4 and Reconnendation 1 The Review Group previously considered both the necessity or benefits of sulfur removal and the capability of the proposed peroxide flushing process for accomplishing sulfur removal. At that time we concluded that sulfur' removal was not essential for the return of the plant to power. All available information indicated that the corrosion had stopped and that sulfur residuer,following completion of the repair would be comparable to other plants. The primary benefit of sulfur removal was intangible; the potential for reactivation of the corrosion from these surface residues would be reduced in proportion to the degree of effectiveness of removal. However, there was (and is) no quantitative measure of the potential for reactivation.
The proposed peroxide ~ flushing process was at a very preliminary stage of development at that time. The Review Group expressed concerns over potential corrosion due to this process (alone and with the presenceofsulfur),giventhatperoxideconcentrationsandexposure times exceeded current industry experience. We also pointed the
,# difficulties of scale-up and the need for better understanding of the i
process. We could not assess whether the proposed process would l
accomplish its objective of sulfur removal without harm to the plant.
Since that time GPU Nuclear has generated additional data showing that sulfur concentrations are an order of magnitude higher than originally reported, that the sulfur exists as a sulfate at the surface of the oxide, and that nickel sulfide exists near the metal surface. They were unable to develop comparative data for other plants, however, so that we still do not know if these observations are unique to TMI-1. GPU Nuclear has also consulted with other experts concerning the desirability or necessity of sulfur removal, and has received
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strong recommendations to remove sulfur.
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GPU Nuclear has also aggressively pursued development of the peroxide flushing process. Additional materials have been added to corrosion 1
test programs to determine if peroxide flushing would have detrimental l
affects on core winterials. None were found. The cause of anomalous data from the beaker test was determined. Loop testing has been performed which has defined the process parameters and documented the perfonnance of the process.
Ion exchange resins have been selected and tested for sulfate and anunonia removal. These results significantly reduce our prior concerns about the peroxide flushing process itself.
The Review Grous, continues to believe, however, that sulfur removal is not essential for safe operation of the plant, and that the costs and residual risks in uncertainty over peroxide flushing outweigh any benefit. We believe that the corrosion process is presently passive and will remain passive with good chemistry control even though sulfur residues-will be available. We note that tests show 20-50%
of the sulfur will not be removed by the process, so that sulfur residues will still be available after the flush. This process will be costly in time, chemicals, ion exchange resins, radioactive waste i
generation and man-Rems.
In any complicated process, upsets can occur which could result in exposure of system materials to conditions not enveloped by testing.
Finally, there is much about the reactions between peroxides and system materials which is not understood, so that (in spite of testing) there remains a risk that the process could be detrimental.
We therefore believe that peroxide flushing to remove sulfur is not essential to plant safety nor is peroxide flushing expected to have an adverse effect on plant safety.
4 Irrespective of our beliefs concerning the necessity for and questionable benefits of sulfur removal, we connend GPU Nuclear's efforts toward l
developing the peroxide flushing process. We have the following l
connents to offer.
Further Connent 1 - Peroxide Flush Process Control - Table B-1 of TR-010, Safety Evaluation of TMI-l Reactor Coolant System Cleaning, presents the anticipated control parameters and and frequency of sampling for the process. The sampling frequency is too low to provide adequate control, especially during initiation and transients such as sulfate removal, residual peroxide / oxygen removal and annonia removal. GPU Nuclear indicated that this Table is considered a minimal frequency, and that the procedure will address these concerns.
Further Comnent 2 - Perfonnance of Sulfate Removal Resin - The sulfate removal resin was tested for efficiency at a pH of 7.9, in contrast to the maximum pH of 8.5 for the process. Higher pHs will be more limiting, possibly resulting in lower sulfate removal efficiencies and chloride throw from the resin.
It is suggested that analyses or testing be performed to evaluate the performance of the resin over the full range of process conditions.
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Further, the potential for chloride throw from the resin should be evaluated and, if necessary, additional analyses performed to monitor chlorides during the process.
Reconsnandation 2 - The GPU Nuclear response is satisfactory.
Reconsnandation 3 - The GPU Nuclear response is satisfactory.
D.
Removal of Sulfur Residues This subject was covered under Section C above.
E.
Stress Analysis of Steam Generators Comments 1 through 4 - No GPU Nuclear response was required.
Consnent 5 - Leak Before Break - The Review Group previously consnented in part "In the transition zone, two additional stress states are superimposed on the axial tensile stresses. One stress state is caused by the axial load, and the other is caused by the interfacial pressure between the expanded tube and the tubesheet".
The sunr of these three states is the stress state in the transition As far as leak before break is concerned, the precise zone.
description of these states is not necessary. But if conclusions, on main steam line break are drawn, they should include the transition zone.
For this reason, the Review Group reconinended that a detailed stress analysis 'of the transition zone be made including the loading of the main steam lic.s break.
Subsequent to the Review Group' meeting, GPU Nuclear advised that they had completed the stress analysis of the. transition zone, and the revising the stress report to include this analysis and that the analysis shows an acceptable striss condition. This resolves the Review Group's comment.
F.
Steam Generatcr Leak Tightness After Repair Reconenendation 1 - GPU Nuclear establishes administrative limits on leakage which consider the threshold level of detectibility. The response resolved the -Review Group's consnent.
i However, the additional analysis of leak before break conducted by l
GPU Nuclear and reported in Safety Evaluation Report 008 raised a further question. Warning before break is an important safety consideration in the operation of TMI-l steam generators.
It is important that flaws
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be detected by inspection or leakage before they can grow to critical sizes which would be unstable under normal operating conditions or accidents such as a main steam line break.
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. l' Usually, eddy current inspection and conservative tube plugging criteria are relied upon in demonstrating integrity for steam generator l
tubes. In addition to taking these steps, GPU Nuclear has performed a fracture mechanics analysis to show that the growth of flaws due to mechanical effects is slow (i.e., growth to a critical size takes many operating cycles) and that eddy current is capable of detecting flaws before they approach critical size.
Using the results of fracture mechanics analysis, GPU Nuclear also calculated leak rates through flaws. Their calculations show large margins in both time and flaw size between detection of leakage and failure of a tube during normal full power operation. The same calculations show that a flaw which is of a critical size for a main steam line break will leak approximately 0.1 gpm during normal full i
power operation.
GPU Nuclear has established an administrative limit of 0.1 gpm prirrary-i to-secondary leakage for shutdown. On the basis of fracture mechanics and leak rate calculations, GPU Nuclear believes that this administrative limit assures leak before break for all normal and abnomal operating conditions including the limiting case of a main steam line break.
Comparative analyses by others show that fracture mechanics and leak rate calculations are sensitive to assumptions such as loads in steam generator tubes, flaw shape, flaw surface roughness and the existence t
of threshold stress intensities for crack growth. Because of these uncertainties in the analyses, the Review Group questioned whether the results of the GPU Nuclear analysis of leak before break had sufficient margin for the limiting case of a main steam line break.
Subsequent to the meeting GPU Nuclear has pursued this issue and advised they have come to the following conclusions:
The GPU Nuclear study on crack preparation and their interpretation of the draft analysis done by others suggests that cracks will not grow or not grow rapidly as a result of flow induced vibration. Although growth rate is a function of the assumed threshold stress intensity, even the extreme case of no threshold revealed long time periods for crack growth to a critical size and therefore ample time for operator action to shut down the reactor prior to a tube failure either at power or during shutdown.
The tube loads associated with the steam main line break case wre calculated using a generic analysis for B&W plants.
The assumptions used in this analysis are very conservative with respect to the particular plant parameters for TMI-l and result in calculated tube loads substantially greater than would actually occur. When more realistic tube loads are taken into i
account, the critical crack size is estimated to be significantly larger and the corresponding leak rate is increased by approximately a factor of two. Thus GPU Nuclear concludes
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< that because significant margin exists in the tube loading used to determine critical crack size, no further conservatisms need be added in designating administrative limits which l
take credit for leak before break.
On the other hand, the Review Group considered that addy current inspection should give adequate warning of flaws which could become If unstable before the next inspection for a main steam line break.
further assurance of leak before break is required for the main steam, line break, then more sensitive leak datit: tion techniques might be applied. GPU Nuclear should consider the following:
l Leak rate measurements during cooldown loads as a sensitive way to detect through wall cracks.
Sensitive secondary-to-primary leakage measurements in steam
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generators during shutdown.
GPU Nuclear agrees that addy current testing will identify cracks that could, before the next inspection, become unstable under main steam line break loading. As previously described the threshold for
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detectability for eddy current testing has been determined using notched calibration standards and laboratory-grown sulfur-induced intergranular cracks. This threshold has been found to be below the crack size that would rupture under main stcam line break loads.
In addition, in response to Review Group suggestions, GPU Nuclear will record the condenser offgas activity data during cooldown and evaluate the feasibility of using this data for detennining primary-to secondary leak rates during conditions of higher tube-to-shell delta T.
GPU Nuclear also plans to use secondary-to-primary bubble testing tube as one technique for locating leaking tubes whenever the plant is shut down in response to an increase to a leak rate 0.1 gpm or more.
The high sensitivity of this measurement technique provides additional assurance that flaws that could become unstable before the next addy current inspection will be detected.
The Review Group considers these actions satisfactory.
G.
Plant Operations Reconnendations 1 and 2 - The GPU responses were satisfactory.
Attachments:
Appendix A - GPU Nuclear letter to Members, TMI-1 OTSG Repair Program Review
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Group;
Subject:
TMI-1 Steam Generator Repair; dated April 7, 1983 (with attachment 1 only)
Appendix B - GPU Nuclear letter to same;
Subject:
same; dated April 4,1983 1
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REPORT OF THIRO PARTY REVIEW OF THREE MILE ISLAND, UNhi 1 STEAM GENERATOR RE' PAIR
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s R. F. Wilson - Vice President, Techi:1 cal Functions
~~GPU NUCLEAR i Prehand'by THIRD PARTY REVIEW GROUP:~
Stephen D. Brown
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Stanley A. Holland 7.
Arturs Kalnins William H. Layman David J. Morgan Richard W. Weeks Edwin J. Wagner - Chairman. -
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Submitted for the Review Grou O "^--
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THREE MILE ISLAND, UNIT 1, STEAM GENERATOR REPAIR b-x CONTENTS PAGE 1
PURPOSE 1
SACKGROUND 4
CONCLUSION FIN 0lNGS, COMMENTS AND RECOMENDATIONS*
J A. - Steam Generator Inspection.
5 8.
Cause of Tube Cracking 7
9 C.
Materials Application D.
Removal of Sulfur Residues 12 L-E'.
Stress Analysis of Steam Generators 14 F.. Steam Generator Leak Tightness A.'ter Repair 18 19 G.
Plant Operations APPENDICES:
A.
Interim Report of Third Party Review of Three Mile Island, Unit 1, Steam Generator Repair B.
Letter, R. F. Wilson, GPU Nuclear, to Third Party Review Members, dated April 12, 1982 C.
Third Party Review of TMI-l OTSG Repair Program - Charter D.
TPR Bibliography E.
Compilation of Review Group Questions & GPU Nuclear Answers l
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1 PURPOSE:
This is the eport of the Third Party Review Group established ' y GPU Nuclear o
to provide a timely, independent and 'objecti.ve safety evaluation of the activities being conducted to repair and return the TMI-1 steam generators to operation.
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An inter"im report of the Review Group was issued September 27, 1982.
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It reported the evaluation of part of the THI-1 Steam Generator Repair Program - that relating ta the safety of conducting the proposed repair of the steam generators while the plant is in a cold shutdown condition.
including the effects of the repair on the steam generators and on the remainder of tha TMI-1 plant. The Interim Report is Appendix A to this report. Appendix A is considered an integral part of this report and its '
content is not repeated here.
BACKGROUND:
' On April 12,.1982, R. F. Wilson of GPU Nuclear established a' Third Party Review of the TMI-1 steam generator repair program. Appendix B is the GPU Nuclear letter that established the Review Group. A Charter was supplied which defined
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the purpose, scope, menbership and operations of the Review Group. During the first
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meeting of the Review Group on April 23, 1982, the Review Group discussed the Charter and suggested changes. The revised Charter dated April 27, 1982 is Appendix C.
The evaluations of the Review Group have been conducted in l
conformance with Appendix C.
s The membership of the Review Group, selected by GPU Nuclear for expertise in the following technical specialties is; e.
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Specialty Name Affiliation Steam Generator E. J. Wagner Burns and Roe, Inc.
I Design -
l Chemistry D. J. Morgan Pennsylvania Power & Light Materials R. W. Weeks Argonne National Lab Stress Analysis A. Kalnins Lehigh University' Safety Analysis W. H. Layman EPRI - NSAC Plant Ope, rations S. A. Holland Duke Power Co.
Non-Destructive S. D. Brown EPRI - NDE Center Examinations E. G. Wallace of GPU Nuclear is a non-voting member who seryes as liairon with GPU Nuclear and was assigned as Secretary. All members have been present and participated in all meetings of the Review Group.
In its April 23, 1982 meeting, the Review Group elected Mr. E. J. Wagner to be Chairman.
As can be seen from the expertise of the membership, GPU Nuclear carefully
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selected the Review Group to obtain a broad, range of. competence in technical specialties important to the steam generator repair. ' In addition to obtaining competence, the formation of the Review Group focusec on independence and objectivity. The individual members are not responsible for the performance of the design or development activities involved in the steam generator repair;
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nor are tK organizations with which t.4y are affiliated perfoming the steam generator repair activities.
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It should be noted that the organizations with'which the members are affiliated have many competent personnel and other resources applicable to the subject of this Review Group. The meir6ers have used these resources to conduct their evaluations.
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The members also availed chemselves of personal contr. cts with other experts in universities and laboratories. Notwithstanding, the members acted as Independent individuals on this Review Group. Neither their individual statements nor their contributions to the. reports of this Review Group are intended to '.
represent the opinions or conclusions of the organizations with which they are
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affiliat'ed or of the other experts they consulted.
1 The evaluatiori was conducted using reviews of pertinent documents, submittal of written questions to GPU Nuclear, written responses by GPU Nuclear, review of i
specialty topics by individual members, presentations by cognizant GPU Nuclear or contractor personnel, Review Group meetings and executive sessions of the Review Group members only. Appendix D is a bibliography of the documents supplied to and used by the Review Group. Appendix E is a compilation of the members' written questions and the GPU Nuclear responses.
Full day meetings of the entire Review Group were held on April 23, May 20 and l
21 August 24 and 25, and December 7, 8 and 9,,1982. At each meeting, GPU Nuclear i
and its contractors presented status and plans, answered oral questions from the Review Group and were given suggestions and conments by the Review Group. The b
consents of the Review Group were summarized with senior GPU Nuclear I
Management at each meeting. These sunnaries were made to Mr. R. F. Wilson, Vice President, Technical Functions in the first three meetings and to both him snd Mr. P. R. Clark, Executive Vice President at the December 9, 1982 meeting.
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L NRC staff menbers attended and participated in' the Rev.iew Group meeting of l
August 25, 1982. GPU Nuclear, contractors and Review Group members answered questions from the.NRC staff members. The summary meeting of I
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4 December 9,1982 was attended by NRC staff members and a representative of the I
State of Pennsylvania. They heard oral presentations of the findings of the t
j Review Group made by each of the members in his field of expertise. Questions fion all attendees were responded.to by Review Group members.
The principal foci of this review were the two Safety Evaluation Reports prepared by GPU Nuclear. The first treated only the repair of the steam generators.
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j identified as item 10.1 in Appendix D.
The Review Group's evaluation was reported"in Appendix A and is not repeated here. The second Safety Evaluation j
Report treated the return of the TMI-l' plant to service after repair of the iteam generators.
It is item 17.'9 of Appendix D.
The following' evaluation addresses this second Safety Evaluation Report and the activities it embodies through December 9,1982.
l CONCLUSION:
The Third Party Review Group recognizes that work has progressed to conclusions 1
in many areas. supporting the evaluation of safety, such as the cause of the steam generstor tube cracking, corrective actions, and actions to prevent recurrence. Other supporting efforts have not yet neared cogletion.
Recognizing the status of all the activities supporting the Safety Evaluation.
i safe operation of the TMI-1 plant after repair of the steam generators will be dependent on several remaining major activities:
1.
Post repair testing and hot functional operation of the systems.
.a 2.
Coinpletion of analyses including leak before break and the s
d contingency of multiple tube rupture.
3.
Translation of analytical 'sork such as leak before break and
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uultiple tube rupture into useable plant guidance, procedures and training.
4.
A conservative approach of power es:alation after completion of repairs.
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GPU Nuclear work in these areas has not reached the point where the results l,
can receive a final safety evaluation.
It would be premature to evaluate that when all existing GPU Nuclear plans are completed that the Third Party Review would conclude that the results will be positive and will ensure that plant operation would be without increased risk.
1 Gpu Nuclear has performed sound technical work and the Third Party Review Group anderses the planned GPU Nuclear program to insure safe operation. We beli, eve that there is a high probability that When the GPU nuclear program embodied I
in the Safety Evaluation Report is completed, that the TMI-1 plant with steam generators repaired in accordance with present plans will be able to operate without i
undue risk.
FINDINGS, C0tHENTS AND RECOMMENDATIONS:
A.
Steam Generator Inspection Finding 1 - GPU Nuclear and their eddy current inspection contractor have done a thorough job of improving and qualifying eddy current inspection techniques for detecting the primary-side cracking condition which exists in the TMI-1
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steam generators. A minimum detectable defect condition has been identified i.
based on inputs from stress analysis and fracture mechanics :onsiderations.
Eddy current standards have been, fabricated which duplicate the desired detection limits and eddy current system optimization studies have been 1
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f conducted which have identified test frequency, coil fill-factor and instrument sensitivity settings. Confirmatory detection studies have also been conducted using intergranular stress corrosion cracking (IGSCC) specimens wfiich were fabricated in the laboratory. Thisqualification program, and the extensive inspection performed on both steam generators with the qualified techniques, provide confidence that the primary side cracks above the defined minimum detectable size have been detected.
L Reconenendation 1 - The Review Group recommends the plugging of all tubes which contain ID fndications between the upper and lower tube sheet in bott-steam generators.
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Eddy current indications have been detected between the upper and lower tube sheet. These indications are a mix of ID and OD and in general have an eddy current estimated depth of less than 40 percent through-wall. GPU Nuclear stated that they plan to leave some or all of these tubes with ID indications in service. About 60 tubes may be involved. The development of a plugging limit requires knowledge of eddy current measurement accuracy, defect growth rate and transient steady / state tube load conditions: Each of these factors has uncertainties which are not known with confidence. Although sufficient operating experience with other once-thru steam generators (OTSGs) would justify allowing the OD indications less than 40 percent through-wall to remain in service, the ID indications
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are most probably stress corrosion cracks and should be plugged.
Reconsnandation 2 - Tubes within three rows of the lane region and in the h
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l wedge-shaped region at the periphery which have OD indications at the 15th support plate or above, should be plugged as has been done in other OTSGs.
Reconenendation 3 - The eddy current inspection baseline performed after expanding tube ends should be extended to the full tube length on a 7
selected number of tubes to detect possible deleterious effects of' the b
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.c explosive expansion. Tubes which contain defects less than 40 percent through-wall mentioned above should be considered as the sample set to detect evidence of defect growth or initiation as a result of the explosive expansion.
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Recommendation 4 - The Review Group recommends selected.non-destructive examination of seal welds and tube ends on the lower tube sheets to confirm L
the absence of any deleterious effects of the explosive expansion of the i-upper tube ends.
B.
Cause of Tube Cracking Finding 1 - The Review Group is in agreement with the failure scenario presented by GPU Nuclear in Section II.D'.2 of the Safety Evaluation Report for return to service Appendix 0, item 17.9.
f The probable mechanism for the tube cracking was IGSCC or possibly stress-assisted intergranular attack (IGA) resulting from exposure of the tube ID G
to sulfur and its lower oxidation states during cold shutdown with the reacter f
coolant system partially drained. The most plausible input corrodants were sodium thiosulfate which probably leaked from the containment spray system into
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the reactor coolant system during the extended shutdown and oxygen which was presented in the gas phase of the partially drained reactor coolant system.
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The concentration of the' damage at the upper tubesheet region,can be u
explained by two mechanisms, both of which may have been present. Sodium thiosulfate or corrosive sulfur compounds from the oxidation of sodium thiosulfate may have been concentrated at the liquid-gas interface which
!i was in the vicinity of the upper tubesheet by evaporation following N-l U
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fluctuations in the reactor coolant level. Alternatively, the solution near the. interface would have had better access to oxygen in 'the gas phase, resulting in a higher concentration of c3rrosive species in the water i.
near the interface, I
i The spread in damage may be the result of fluctuations of reactor coolant 4
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1evel and/or the concentration of oxygen in the solution. The oxygen concentration would be expected to decrease with increasing distance from the interface.
Evidence supporting this scenario is as follows:
4.
Destructive examination showed that the : racking was intergranular, ID initiated, and circumferential1y oriented. This indicates a strong tensile stress component on the cracking mechanism. Since major tensile stresses in the axial direction are expected only during cool-down and cold shutdown, this implies that the cracking occurred under these conditions, b.
Areas of IGA were found on tube samples from the gen'erator, both on the ID surface and in some cases on metal bordering the fracture surfaces.
i As GPU Nuclear pointed out, the surface IGA may have been generated by pickling during tube fabrication.
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c.
The tubing was sensitized, a condition known to be susceptible to IG5CC when exposed to the lower valence sulfur oxides (polythionic acids).
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I-These tend to be stable only at low temperatures, further evidence l
l that the cracking may have occurred at low temperature.
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d.
Cracking has been induced by GPU Nuclear on stressed Inconel-600 Pd specimens exposed to sodium thiosulfate and oxygen at low temperatures.
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t e.
The damage was widespread and extensive on both generators. This implies that either relatively large quantities of corrodantssor a highly corrosive species would be required. Sodium thiosulfate and oxygen appeared to be the only known corrodants which may have been present in sufficient quantities to produce the extent of attack cbserved.
Reconnendation 1 - Although the Review Group believes that reduced sulfur forms were the most likely corrodant, we recommend that GPU Nuclear implement corrective measures or verify their existing programs for minimizing ingress of all impurities (not just sulfur) into the reactor coolant system.
For' example', failure analyses have consistently reported carbon as the major i
impurity on ' fracture surfaces, followed by sulfur. Carbonates in the presence of oxidants at high temperature can produce IGA and IGSCC of Inconel-600.
Other contaminants (lead, mercury, phosphorus) can also induce IGSCC.
Specific areas where attention is recommended are as follows:
a.
Resin loss from the purification ion excharigers, since thermal de-l composition of cation resins can release ulf::r acids.
i b.
Contamination of the reactor coolant system makeup by greases, oils, organic solvents, or possibly resin fines from the makeup demineralizers or from' water recovered from liquid radioactive waste treatment.
c.
Contamination of the reactor coolant system or its makeup from peripheral, connected systems such as sulfur in the waste gas vent collection system.
l C.
Materials Application Finding 1 - GPU Nuclear.is to be comended for their materials work to date associate with the steam generator repairs. They had two independent metallurgical failure analyses perforried in addition to their own efforts, and they were diligent in resolving minor differences in the findings. They assembled
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a separate panel of corrosion experts to advise them, and have pursued l
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their advice by inspecting other areas of the reactor coolant systems j.
considered especially susceptible to similar attack by establishing a test i
, rogram to determine the feasibility of cleaning the residual sulfur out of p
the system, and by establishing short and long-term corrosion testing programs to help in identifying any' residual corrosion problems prior to their
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possible occurrence in the plant.
In general, they are doing a thorough job in attempting to assure the future relf ability of the materials in the
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system.
In reviewing the foregoing efforts by GPU Nuclear and their cartractors, the Review Grous expressed the following notes of caution:
Coment 1 - Although nondestructive and limited destructive tests were carried out in looking for possible stress corrosion cracking in the rest of the reactor coolant system and none was found, such cracks tend to be i
very tight and are indeed very difficult to detect. Yet all the ingredients to generate such cracks were apparently present; i.e., sensitized and stressed susceptib1e materials (due to welds), and presumably a thiosulfate-contaminated I
aqueous environment. Therefore, GPU Nuclear should remain alert to the possibility that small cracks may, in fact, be present in susceptible components of the reactor coolant.
l, Comnent 2 - Through a fracture mechanics analysis, GPU Nuclear arrived at a tentative conclusion that steam generator tubing defects below a certain.
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size range will not propagate due to flow-induced vibrations. The analysis l
which led to this conclusion depends on a large extrapolation of a s
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limited crack-propagation-rate data base. This makes it hard to.
substantiate a firm conclusion.
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Coment 3 - The long-term corrosion tests, which are designed to anticipate problems before their possible occurrence in the plant, do include most of f.-
the right ingredients and should be very helpful. However, they do I
not include a flow-induced vibration type of loading which could make a significant non-conservative difference in,the results once a crack is initiated..
e-
- Coment 4 - Cleaning of the residual sulfur in the system poses a dilema g
since even the laboratory-scale beaker test results are apparently not fully understood at this time. Some of the test results have shown erratic
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cleaning and peroxide consumption. The time required to remove sulfur is greater than. expected, and the peroxide concpntrations are much greater than previously used to encourage crud removal in other nuclear plants.
In recognition of these uncertainties, the Review Group recomends the' following.
Recommendation 1 - If GPU Nuclear pursues development of the peroxide process for sulfur removal, the effect of greater than 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> exposure of core materials (e.g. Zircaloy) to hyc' ogen peroxide at the anticipated r
i concentration and pH conditions should be included in the test program.
Also, scaled-up tests should be done in metal systems at least somewhat more closely simul,ating the reactor cooling system environment.
Recomendation 2 - To gain experience in operating the unit while keeping I
the risks as low as possible,'GPU Nuclear-shouTd consider substantially extended operation at low power during a slow and deliberate power escalation the first time the plant goes critical. Although we do not have an l
analytical basis for a specific tiuration, a hold period of perhaps a month i :._
or more at 40 percent power should be considered before the Loss of r-n-
t
- 12,
7 Feedwater/ Turbine Trip test is performed. This might be followed by
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.another month or more at 70 percent power before final escalation to 100 percent power. Also, thi[s first power operation might better be terminated
,by a normal cooldown procedure rather than by the Overcooling Control Test which is currently planned. This last test could be done during subsequent opera'tions.
. Recomendation 3 - GPU Nuclear should c'onsider the possibility of deliberately
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running one steam generator at a higher power than the other during the first powe'r escalation hold periods. The objective of this would.be to force..any possible operating problems to occur in the higher power steam generator before such problems affect both units. We understand a sub-stantial power unbalance between loops is within the range of plant design.
We recognize, however, that this recommendation may involve other operating considerations which would have to be weighed before a decision could be made.
D.
Removal of Sulfur Residues Finding 1 - GPU Nuclear has proposed that sulfur residues known to be present on reactor coolant system surfaces be removed by a hydrogen peroxide flushing process to eliminate the possibility that these could produce future attack. The Review Group recognizes the logic that, if sulfur caused the problem, it is conservative to remove it under controlled conditions rather than possibly letting it come off under uncontrolled conditions.
However, the Review Group is not cony'inced that this sulfur removal proces.s will be of much benefit.
Further, we do not think it is essential for the return of the plant to power. This is based on the following considerations:
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a.
Eddy current testing indicates no progression of cracking since it was initially observed.
b.
Although the data base is very limited, it appears that present sulfur concentrations on THI-1 surfaces are comparable to those at other plantsofthistype.
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c.
The repair activities being conducted in the OTSG will result in substanthal flushing and wiping of the tube surfaces (the major portion of the reactor coolant system surface area) as well as the upper head of the OTSG. These are the regions of maximum' expected surface sulfur concentrations.
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Finding'2-The sulfur removal process has,not been developed to,the point that the Review Group can assess whether it will accomplish its objectives.
Review of the available information raised the following concerns:
a.
Although beaker tests have shown that peroxide additions can oxidize and solubilize sulfur, scale-up from beaker testing to predict conditions under which this will work on the plant may be difficult.
In contrast to the glass beaker, the metallic surfaces.of the actual plant with their burdens of corrosion products could be expected to cause different process dynamics and results.
b.
The beaker tests have shown some anomalies on peroxide concentratien change and sulfate response which indicate the need-for more under-standing of the reactions occurring.
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c.
The process as currently envisioned will be protracted (about 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br />) and will require significant cleanup efforts to remove the large concentrations of ammonia required for pH adjustment of the reactor coolant *.
d.
The concentration of peroxide envisioned (about 25 ppm H 0 ) is 22 significantly higher than that typically used in other nuclear plants to enhance crud removal. This higher concentration may require a more careful assessment of materials compatibility than presently given.
especi, ally in light of the long flush times expected.
e.
These flushes are known to put significant amounts (1-2 ppm) of nickel tiitosolutionandtocausecrudbursts. These effects should be considered.
f.
Fina'ly, although the corrosion testing program is considering this, there is concern that peroxide flushing could produce the corrosion
' 'that we are trying to prevent.
Recommendations are included in Section C. Materials Application of this report that take cognizance of these findings.
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E.
Stress Analysis of Steam Generators Finding 1'- The main question that the Review Group addressed in this area is the following: did the events that occurred after the initial shutdown, including the repair, affect the OTSG tubes of TMI-1 in such a way that, during restart and subsequent operation, the stress levels can be expected to be significantly higher, or the strength of the tubes significantly lower, than those in a normal GTSG*l From the infonnation received, the answer appears to be negative, with a minor qualification regarding the undetected defects that are left in the tubes (see Coment 4 below). Therefore, the m
Review. Group concludes that the integrity of the tubes in the OTSGs of e-the TMI-1 has not been significantly reduced to influence the restart and subsequent operation of the plant. This conclus' ion is based on the following
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Coment 1 - Effect of cooldowns and cold shutdowns on strength of tubes -
The Review Group has found no evidence to suspect that the cooldowns, starting-from the one in April 1979, have subjected the tubes to stresses that are higher than the design stresses. During the two cold shutdown periods, the one after April 1979 and the other after September 1981, the tubes have been in a state of tensile stress, although, from the information received, the levels of these stresses could not be detennined with any accuracy. An indication of the stresses comes from another plant in which the gap between ends of a broken tube translated to a tensile stress of about 4000 psi, well below the allowable stress.
Since creep at the cold shutdown temperatures and stresses should be insignificant, the Review Group concludes that the cooldowns and cold shutdowns have not left the tubes in a w aker state than in any other OTSG.
Coment 2 - Effect of repair on strength of tubes - The explosive expansion of the tubes could affect the stress levels, if the process would change the strength or some dimensions of the tubes. From the information that i
the Review Group has received, from the reports on the qualification tests, and from the statements made in publications issued by the tube expansion contractor, the Review Group' concludes that the repair process is not expected to affect significantly the stress levels in the tubes in the restart and subsequent operation periods.
Coment 3 - Effect of environment on strength of tubes - Since the Review Group has found no indication of significantly higher stress levels than in normal OTSG tubes, it concludes that a corrosive environment, and not i
abnormal stress levels, must have been responsible for the appearance of the cracks.
It concludes also that, if the environment is made more favorable, then, at the same stress levels, cracks should not propagate L.,
in the OTSG tubes if they were in the same condition as in any other OTSG.
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f Comment 4 - Effect of defects on strength of tubes - The Review Group recognizes that, at this time, the tubes probably have some small defects that were not detected by the eddy current tests and 'were not eliminated by the repai'r. These defects present the potential for leaving the tubes in a weaker condition than those in a normal OTSG.
Both GPU Nuclear and the Review Group have addressed the question of what could happen to these small defects during the restart and operation.
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GPU Nuclear.has performed an extensive analysis of the possibility that these small defects could propagate as fatigue cracks if the tubes were subjec5edtoflowinducedvibration. This analysis concludes that the cracks will grow at a stable growth rate within the tube wall, and that the time that is required for a crack to reach the OD of the tube is longer than the lifetime of the OTSG. GPU Nuclear has calculated also the leakage rates through through-cracks and concluded that they are high enough, so that leaking tubes can be detected and taken out of service, before the cracks become unstable.
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The Review Group finds.these results' reassuring, but has some reservations with regard to the limited data base for the crack propagation rates that has been used in the analysis. Coment 2 on Materials Application refers to this uncertainty. The Review Group recognizes that bett'er data are not available at this time and acknowledges the difficulty of making firm conclusions in this case on fatigue-crack growth rates and their stability.'
l" The Review Group's reservation on this matter is somewhat mitigated by 1
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l observing that the accuracy of the predicted growth rates, even though important, may not be as crucial for the present purpose as the analysis of the manner in which the cracks may grow. The Review Group recognizes q
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17 the possibility that among the-undetected defects there may be some that are large enough to break through to the OD and propagate along the circumference in a stable manner, with the potential of breaking the tube when the crack becomes urtstable. Such defects may have simply escaped detection and are the primary candidates for breaking some tubes. The crucial question i
is whether or not these tubes will leak before they break, so that they can be taken out of service before they do some damage.
Conment 5
'.eak-before-break - In the event that a defect does grow in the form of a crack through the wall, and eventually breaks the tube, the Review Group has addressed the following question: Will the fracture process ensure Leak-before-break? TheReviewGroupconcludesthatthe answer is positive and offers the following arguments.
EPRI Report NP-2399, dat.ed May 1982, states that small defects at the ID of a tube have about the same fatigue-crack growth rates toward the OD as along the, circumference, provided that the stresses throughout the wall are axisymmetric and tenstle. Since in the free span (away from tube sheets),
the relevant stresses in the tubes are axisynmetric and tensile throughout the wall, then,a fatigue crack in the free. span will break through the wall and produce leakage before it grows around the circumference and breaks the tube, thus ensuring Leak-before-break.
(Note that the question
- of threshold detectability is treated in Section F. Steam Generator' Tightness After Repair.)
However, in the expansion transition zone of the tube, in the vicinity where the expanded tube diameter changes to the nominal diameter, two additional stress states are superimposed on the axial tenslie stresses.
One stress state is caused by the bending stresses in the transition zone, that are produced by the axial load, and the other is caused by the interfacial pressure between the expanded tube and the tube sheet. GPU Nuclear has performed calculations of these stress states, and they show a
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I rapid decay from the transition point. This means that the transition zone lies well within the tube sheet. The important. point is that these stresses introduce compre,ssion within the tube wall in the transition zene, o
which, according to EPRI NP-2399, can make fatigue cracks grow faster
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along the circumference than toward the 00. This means that in the transition zone the tube could b'reak before it leaks. However, since
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the break would occur within the tube sheet, the end of the broken tube would be restrained in the hole, and a controlled leak would result.
Based on this argument, the Review Group has concluded that such leaking tubes, broken in the trans,ition zone, could be detected and removed from service' before an excessive react'or coolant leak rate results.
F.
Steam Generator Leak Tightness After Repair Finding 1 - The Review Group expects that operational leak rates, principally
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from leakag'e through the explosively expanded tube joints, may exceed the process qualification requirements.
T'he steam generator repairs have been qualified to requirements for a
- I-maximum total leak rate of 1 lb/hr. The qualification program has demon-strated the capability of the process to meet this requirement. However, industry experience with high quality, expanded, unwelded tube-to-tubesheet
[,
joints would. indicate a low probability that this leak tightness will be obtained in these steam generators. Each repaired steam generator will l1 depend upon several thousand expanded tube joints. Even if this leak tiga'-
g' ness is initially attaint,d.. leakage wi.11 tend to increase with future operation.
GPU Nuclear has further evaluated the effects of operation with a leak i
rate through the steam generators of 6 sal /hr. (about 50 times the 4. -
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qualification maxio m rate). Operaticn was found acceptable by"GPU Nuclear. Plant modifications are in prugress to facilitate operation with F
leak rates in this range and administrative controls will be imposed directly on leak rates, or on related radiological conditions. We recognized i
that GPU Nuclear has evaluated and prepared for this eventuality.
Reconnendation 1 - The Leak-before-break evaluation should include con-sideration of a realistically high steam generator leak rate and this rate should be consistent with the administratively imposed controls on oper'ation with leakage through the steam generators.
GPU Nuclear did not explicitly treat this operational leak rate in the Leak-before-bre'ak evaluation. This evaluation is sensitive to the threshold detecta'bility of a leak from a tube defect. This threshold is in turn dependent upon the total leak rate from all sources through the steam generators during operation.
G.
Plant Operations l
Finding 1 - GPU Nuclear has satisfactorily addressed 'most of the operational i
considerations of concern to the Review Group. Others, particularly those relating to Leak-before-break and multiple tube failures are still in development and their adequacy cannot be adjudged.
i Specific concerns that have been addressed include:
That normal operating parameters will need to be changed as a result of the a.
l-OTSG repair program. GPU Nuclear has analyzed the plugging effects and has concluded that no significant operating parameter changes will need to be effected for normal operation.
b.
That adequate procedures for abnormal operation are developed and ope.rators are trained with them. GPU Nuclear is in the process of upgrading
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procedures. Operating training is planned, following this upgrade. Multiple tube leak' procedures are included in this upgrade. Operator training under-steam generator leak conditions was completed following the GINNA tube rupture incident. Annual requalification (both class room and simulator) i
' training either has been completed or is presently scheduled.
c.
That,the plant will have capability for handling leakage into secondary systems in event of a steam generator tube leak. Various station modifications are presently being installed to address this concern.
1.
Turbine building sung will be fiberglass lined to facilitate decontamination.
I' 2.
Aiiditionalprocessingequipmentanda 250,000 gal, storage tank I
will be installed to provide reaction time and cleanup capatility.
3.
Va'rious drains will be re-routed t5 reduce we.ter volume collected in the turbine building sump.
4.
Various drains that could become contam'inated will be re-routed l
to the turb ne building sump.
5.
A system will be installed to collect and dispose of condensate polisher resin as radwaste.
d.
That adequate steam generator leak detection capability will be provided..
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Systems with the capability to identify and quancify steam generator leakage will be in place along with appropriate guidance to the operater.
L The following operational considerations could be improved and are the subjects of reconnendations:
Reconnendation 1 - Inspection of the waste gas system vent header branch l
piping identified a pipe cracking problem in the heat affected zone of butt welds. The failures were attributed to sulfur contamination. From a l
safety and operational point of view we recommend this inspection be l
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expanded to include the waste gas decay tanks and associated isolation l
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valves. Reactor building isolation valves associated with this system should also be included'.
Recomendation 2 - We reconnend that during the " slow" approach to power escalation after repairs (Reconnendation C.2 above)", a planned program,
of operator training in the plant be conducted.
We feel this will allow the operator time to feel comfortable and "get I
his arms around" the job responsibility. TMI-1 operating personnel have F
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not been exposed to actual operating plant experience in over three years.
During this period people have been promoted and transferred and now hold b
positions in which they have no real operating plant e.xperience.
Plant modifications have been installed during this period with no opportunity for the operators to gain " hands-on" experience in the use of this equipment.
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