ML20083L842
| ML20083L842 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/27/1982 |
| From: | Brown S, Holland S, Wagner E GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20083L846 | List: |
| References | |
| NUDOCS 8211040575 | |
| Download: ML20083L842 (47) | |
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APPENDIX A INTERIM REPORT OF THIRD PARTY REVIEW OF THREE MILE ISLAND, UNIT 1, STEAM GENERATOR REPAIR i
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To R. F. Wilson - Vice President, Technical Functions
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GPU Nuclear f-Prepared by THIRD PARTY REVIEW GROUP:
Stephen D.
Brown Stanley A. Holland u
Arturs Kalnins Willian H. Layman
'l David J. Morgan Richard W. Weeks Edwin J. Wagner - Chairman 1-L pppyjos Been'Se'ni(6 PDR I
L Submitted for the Review Group by:
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date:
9/27/82
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e PURPOSE:
This is an interim report of the Third Party Review Group to evaluate a part of the TMI-l Steam Generator Rep' air Program -
that relating to 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 the TMI-l plant.
This 1
interim report was requested by R. F. Wilson, GPU Nuclear, i
on August 10,.1982 to obtain the Review Group's evaluation of this part of the steam generator program concurrent with decision making on conducting the repair in TMI-1.
As GPU i
Nuclear completes the remainder of the overall repair program, the Review Group will report its evaluation of the remainder of Scope of Review defined in the Charter for this Third I
Party Review.
t CONCLUSION:
Based upon the information developed by the repair program and summarized in the Safety Evaluation, the Review Group concludes that the proposed repair conducted on the TMI-l steam generators in conformance with the control systems described will not have adverse effects on the nuclear safety related items of the i
plant (including the steam generators) in the cold shutdown condition.
This includes consideration of potential hazards from:
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Missiles generated by the explosive process.
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Introduction of chemical residues in the steam generators, reactor coolant system or the reactor plant ambient.
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Transmission of pressure pulses through air or structures to sensitive items such as previously expanded tubes, previously plugged tubes, other steam generator structures or safety related instruments and controls.
4.
Bandling of axplosives in nuclear safety related equip-ment or structures (We note that no Review Group member is expert in handling explosives.
Acceptance of the response to this potential hazard is based on the procedure controls described in the Safety Evaluation, their compliance with the Laws of the State of Pennsylvania i
1 and the exclusive use of blasters licensed in accordance l
with that Law).
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Occupational radiation exposures.
This conclurion applies only to 'the safety of the plant, including the steam generators, of conducting the proposed repair while the 1
plant is in the cold shutdown condition.
This conclusion does not apply to the safety of returning the plant, with steam generators repaired by the proposed process, to service.
Although efficacy of the repair is not a consideration in the safety of conducting the repair, it will be important to the safety of T-returning the plant service. 'GPU Nuclear may elect to proceed
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with the proposed repair at substantial economic cost.
The Review Group therefore considers it appropriate to render an opinion now about the efficacy of the repair as it may affect future safety considerations.
The Review Group believes that the proposed repair, after completion r
of the on-going qualification and when conducted in accordance with l
the control procedures, has a high probability of producing tube-to-tubesheet joints adequate for a safe operation of the plant.
However, based upon industrial experience with expanded, unwalded i j tube-to-tubesheet joints in high pressure heat exchangers, it is expected that greater leak ratas will occur during normal operation than typically experienced on new nuclear plant steam generators.
The Safety Evaluation covering return of the plant to service should I
consider this possibility APPROACH:
On April 12, 1982, R. F. Wilson of GPU Nuclear established a Third Party Review of the TMI-l steam generator repair program.
A Charter was supplied which defined the purpose, scope, member-L ship and operations of the Review Group.
The evaluations of the Review Group have been conducted in conformance with the Charter.
The membership of the Review Group, selected by GPU Nuclear for expertise in the following technical specialties is:
Specialty Name Affiliation Steam Generator E. J. Wagner Burns and Roe, Inc.
i Design Chemistry D. J. Morgan Pennsylvania Power & Light' Materials R. W. Weeks Argonne National Lab
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Stress Analysis A. Kalnins Lehigh University Safety Analysis W. H. Layman EPRI - NSAC Plant Operations S. A. Holland Duke Pcwer Co.
Non-Destructive S. D. Brown EPRI - NDE Center Examinations
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7 E. G. Wallace of GPU Nuclear is a non-voting member who serves '
as liaison with GPU Nuclear and was assigned as Secretary.
All members have been present and participated in all meetings of the Review droup.
It should be noted that the members act as independent individuals on this Review Group.
Neither their individual statements nor their contributions to any reports of this Review Group are intended to represent the opinions or conclusions of the organizations with which they are affiliated.
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The evaluation reported in this interim report was conducted concurrently with evaluation of the full Scope of Review defined in the Charter.
The evaluation was conducted using reviews of I-pertinent documents, submittal of written questions to GPU Nuclear, written responses by GPU Nuclear, review of specialty l-topics by individual members, presentation by cognizant GPU Nuclear or. contractor personnel, Review Group meetings and l,,
Executive Sessions of the Review Group members only.
Full day meetings of the entire Review Group were held on April 23, May 20 and 21, and August 24 and 25, 1982.
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The focus of the avaluation of this interim report was the Safety Evaluation of the TMI-l steam generator repair distributed to the Review Group (E. G. Wallace letter of August 20, 1982 and of its reference documents 1, 2 and 4).
The proposed repair described in this Safety Evaluation is the explosive expansion of the.;p 17 or 22 inches of the tubes within the top tubesheets of both steam generators.
The repair will be made on all tubes which will be L-returned to service.
The explosive expansion creates new pressure boundary joints between the reactor coolant and steam-side of the steam generators.
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COMMENTS:
During the review of the Safety Evaluation and supporting documents, certain observations were made by the Review Group.
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comments and the GPU Nuclear responses could be pertinent to the L-conclusions of the Review Group in assessing the return of the plant to service.
They are therefore documented as follows:
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The Safety Evaluation states that the repair joints will be leak tight and meet the design ' bases of the original joints.
As the Review Group stated under Conclusion, the repaired joints will probably be adequately leak tight for safe operation.
However, the joints should not be expected to be as leak tight in normal r
I operation as those of typical new steam generators.
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The Safety Evaluation for the return of the plant to service should consider the potential for higher leak rates from reactor coolant to steam systems and the l '
handling of resultant radioactivity dis' charges such as from the condenser air ejectors.
GPU Nuclear agreed.
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Although indirect measurements provided some indication, the repair process qualification does not contain a direct metallographic examination to verify that the metallurgical structure of the tube material in the expanded region is not degraded by the expansion.
L GPU Nuclear said that such a metallographic examination would be included in the qualification.
3.
Paragrapn 6.1 of the Safety Evaluation discusses residues left on the steam generator surfaces by explosive expansion.
The Review Group understands that
'j the testing at Mt. Vernon showed greater amounts of residue than expected based upon mock-up tests.
Some cleaning is now expected to be necessary.
GPU Nuclear further advised that a material called Immunol is under consideration as a coating to facilitate ramo'al of v
residues.
It would be applied to the tube surfaces before the explosive expanding.
L The Review Group suggested that specific li=its and' appropriate check methods be included in procedures to preclude existence of detrimental contamination from i
l either the explosive residues or Lumunal after completion
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GPU Nuclear agreed.
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Paragraph 6.5 of the Safety Evaluation indicates that the l
steam generators will be isolated by temporary plugs i
from the reactor coolant system during repair.
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a cognizant contractor person stated that, based upon the Mt. Vernon testing, temporary plugs might not be used.
The Review Group considered it important to assure that
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contaminants from explosives not be allowed to travel into the reactor coolant system.
GPU Nuclear agreed and reiterated that the temporary plugs will be used.
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Paragraph 8.0 of the Safety Evaluation discusses the quality assurance and quality control for the repair.
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The Review Group asked whether a quality plan existed l
specifically for the repair.
GPU Nuclear advised that
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the quality actions are an integral part of each procedure i
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s and that all the repair activities would'be conducted
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in accordance with written"proceduxes anc the TMI-l T
Quality Assurance Plan.
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charge condition, correctness of assembly, etc.) was suggested by'the Review Group as the type of check that,
might be identified by a specific quality plan for the repair.
The intent of such a check would be to detect
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any repair process failure early rather than during testing at the conpletion of repair.
This specific check s
I might be conducted by actually explosively expanding a test joint periodically with production explosives and equipment.
GPU Nuclear advised they would assure that the integral quality provisions of their procedures constitute an adequate quality plan for the repair and that an overall. quality control check such.as suggested would be included.
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APPENDIX B r
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OPU Nuclear M p;.
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100 Interpace Parkway f'
Parsicoany, New Jersey 07054 201 263-6500 TELEX 136-482 Wnter's Direct Oial Number:
April 12, 1982 s
Ifenbers, DfI-1 OTSG Repair Program Review Group
SUBJECT:
IMI-1 Steam Generator Repair Third Party Reviev I'
The corrosion problem that has developed in the TMI-1 stesm k.,
' generators represents a first-of-a-kind condition for commercial reactors.
s., As in any new problem, the opportunity for overlooking important l'
relationships between elements of the problem is greater, therefore, i
additional care is required to assure that all important links are idescified and dealt with. It is for that reason GPU Nuclear felt that an independent third party review effort would be warranted to prov,ide i
additional assurance that the failure mechanisms and repair proposal are
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compatible and appropriate, and do not represent a safety concern when the pis.st returns to power.
To achieve the desired result, GPUN Management developed a charter
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for tha review which I have attached for your information and guidance.
I The membership of the group was developed to bring a broad range of i,
individuals with the necessary expertise related to the steam generator failure together for the review func{ ion. Where available I have provided copies of resumes of the members so that you may become familiar with your fellow members' capabilities. I think that you will agree that we have been fortunate in bringing together the quality people needed to review this unique problem.
It is envisioned that a minimum of two meetings will be necessary to complete the review. The first meeting has been scheduled for April 23 at the GPU Headquarters, Parsippany, New Jersey.
For. your assistance, I have included a simplified map of the routes from major highways and u
airports._ If you would like some assistance with local accommodations, please let me know. The first meeting will begin at 10:00 a.m. and will cover a number of organizational details first. The technical presentation will start about 11:00 a.m. and be scheduled to last until 4:00 p.m. so that people may make evening travel connections.
Subsequent meetings will be scheduled on the 23rd with the final meeting being near the and of May.
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l GPU Nuclear is a part of the General Public Utilities System
In preparation for the first meeting, I have enclosed a number of documents which will provide you with a general background of the program and progress as things now stand.
Your review of these documents will undoubtedly raise questions that you would like addressed at some point.
So to try and aske the first meeting productive as possibia, I would request that yca send or call in questions in advance of the meeting and they will be factored into the agenda. 'the base agenda' that is envisioned will be the same as the one used.in the April 7th NRC sestua meeting. The I
esterial from that meeting is in one of the attachments, and will be updated for the 23rd.
I hope that this package will answer a number of your questions and get the review off to a good start, If you have any further questions, please call gd Wallace at (201) 299-2191, who has been assigned as
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Secretary of the group.
Very truly y'ours, 1
e.ca.4Jataa
. F. Wilson i
Vice President - Technical Functions i"
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LIST OF ATTACHMENTS 1.
Charter, Third Party Review of TMI-l OTSG Repair Program 2.
Membership List and Resumes 3.
Map of Routes to GPU Headquarters 4.
OrSG Repair Task Force Organization 5.
Preliminary Report - Failure Analysis 6.
Preliminary Report - Eddy Current Examination Program 7.
Agenda (Proposed) for April 23rd Meeting 8.
Handouts from April 7 and January 25 Status Meetings with NRC r
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ATTACHMENT 1 THIRD PARTY REVIEW OF TMI-1 OISC REPAIR PROGRAM CHARTER I.
PURPOSE l
It is the intended purpose of this " Third Party Review" (TPR) to provide i
a timely, independent, objective, safety evaluation of all activities d e fined in this charter for conformance cc:
- 1) the NRC rules &
I regulations governing the operation of TMI-1; and 2) the adequacy of the steam generator repair program *that will allow safe operation of the nuclear unit.
II.
SCOPE The scope of this review is generally limited to activities associated with the identification of failure mechanisms and repairs of the TMI-1 Once Through Steam Generators (OTSG's).
The specific task areas to be reviewed are described in more detail in Section IV of this charter.
It is the intent of CPUM Manageant to fully develop and implement repairs to the THI-1 OISG's within the provisions of 10CTR50.59.
It is expected that the TPR will promptly notify GPUN Management of any circumstance,
not already ides 3ified by GPUN, that fails to meet these standards.
III.
IEMBERSHIP t
The membership of the TPR body shall include individuals with expertise in the following specialty areas:
i A.
Steam Generator Design and Performance
- 1 member B.
Chemistry
- 1 member C.
Materials
- 1 member D.
Stress Analysis
- 1 member E.
Safety Analysis
- 1 member P.
Plant Operations
- 1 seaber U
G.
- 1 member l
The TPR shall have a GPUN individual assi;b;d as Secretary for the review.
He will be the general interf ace f.or the TPR membership and GPUN. A Chairman will be elected and report the results of the review directly to the Vice President - Technical Functions in this assignment.
The Secretary will arrange for all review material, meetings and recordkeeping for the team.
Any specific information requests within the
- scope of the TPR should be to the Secretary. The Secretary will be a non-voting member in any matters of the TPR seeking concensus opinions.
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Members of the'TPR will either be from outside the CPUN organizatiotr or from the portion of GPUN not responsible for the steam generator repair or TMI-1 operations.
IV.
SPECIPIC REVIEW AREAS A.
Failure Analysis Program This program is intended to identify the cause of tube cracking and means to arrest it.
B.
Eddy Current Examination Progras This program is to develop and implement an eddy current examination method to identify the ex' tent of the tube cracking probles.
C.
OTSG Performance Evaluation It la the object of this effort to evaluate the impact of the repair procedures on the performance of the steam generators, especially in the area of safety analysis.
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D.
Repair Criteria This program is intended to provide guidance concerning the type of repair to be done on the damaged tubes.
E.
Or$G Repair Program E'
This program covers P.he actual repair of the steam generators.
V.
MEETING FRTQUENCY The TPR shall initially meet to receive a presentation by GPUN and its consultants on the current status and direction of the OTSG Repair Program. At that meeting, they will be presented with initial reference estarial that will enable thsa to assess products then available.
A second meeting will be scheduled by GPUN to make a final presentation to the TPR prior to initiation of the final production repairs.
Other,sestings may be held at the discretion of the TPR.
l VI.
EECORDS The Secretary shall usintain records of all meetings of the TPR. The Secretary shall also maintain a record of all documents reviewed by the membership.
Portions of the material may be proprietary in nature.
Appropriate arrangements shall be made to protect proprietary
.L information when it is used.
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Transcripts may be taken of the final review meeting and used as part of the documentation package available to the NRC in support of the OTSG Espair Program.
A final raport 'will be prepared by the review team which summarizes its findings and conclusions regarding the safety adequacy of the repair program.
The report should make an explicit finding that the approach proposed by GPUN is adequate if that is the conclusion of the review.
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ATTACHMENT 2 ee T-MIMBERS. TMI-1 Third Party Review o
STEAM GENZIATOR DESIGN AND PERFORMANCE ED. J. WAGNER - Director, Engineering and Design, Breeder Reactor Division, Burns and Roe, Inc.
o GIMISTRY DAVID J. MCRGAN - Plant Systems Analysis Section, P.P.& L.
o MATERIALS DR. R. W. WEEES - Associate Director Materials Science Division, Argonne National Laboratory i
o STRESS ANALYSIS DR. ARTURS EALVINS - Professor of Mechanics, Lehigh University o
SAFETY ANALYSIS WILLIAM LAYNAN - Electric Power Research Institute o
PLANT OPERATIONS I
STAN HOLLAND - System Production Engineer - Duke Power Co.
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NON-DESTRUCTIVE EXAMA' NATION STEVE BROWN - Electric Power Research Institute o
SECRETARY ED WALLACE - Manager, PWR Licensing, GPU Nuclear l.
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E. J. WAGNER Education:
BSME Carnegie-Mellon University Post Graduate Education at Case Western Reserve, I
George Washington University and Ohio State University.
1 Current Position:
Director Engineering and Design
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Breeder Reactor Division Burns and Roe, Inc.
Current Additional Assignments:
(1)
Member, EPRI Steam Generator Owners Group Architect / Engineer Advisory Committee (2)
Member, Three Mile Island Unit #2 Technical Assistance and Advisory Group Historical Technical
Background:
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Deputy Director for Engineering, Burns and Roe, Inc.; Deputy Director Technical 1
Evaluation - 1975-1981 L
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Division Engine / ring Manager, Westinghouse Electric Corporation - 1970-1975 Chief, Nuclear Components Branch, Naval Nuclear Propulsion Program (USAEC @SDOE]
and US Navy); Senior Naval and AEC Manager of Numerous Steam Generator Development -
Programs - 1955-1970 Test Engineer, Babcock and Wilcox Company -
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Development of Steam-Water Separators l
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DAVID J~ MORGAN l
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EXPERTISE: Coolant technology of pressurized water power reactors ; water chemistry theory, radio-chemistry, cerrosion control, and chemical engineering analysis of reactor coolant and steam generator chemis try control programs, systems, and equipment.
t PENNSYLVANIA POWER & LIGHT, NUCLEAR PI. ANT ENGINEERING Nuclear _ Analyst, Plant Systems Analysis Provide technical direction to Systems Engineering Group, which is responsible for evaluations of the adequacy of Susquehanna Steam Electric Station (SSES) for scenarios involving failures of equipment or operations, for changes to the plant or for questions relating to the adequacy of the plant as designed.
COMBUSTION ENGINEERING, NUCLEAR POWER SYSTEMS 4/81-7/81: Senior Consulting Engineer, Systems Chemistry
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Responsible for incorporating plant performance observations, research and development test results into water chemistry tpacifications and control programs.
Provided consulting support to operating plants.
6/77-4/81: Supervisor, Coo lant Technology Group i
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Directed up to five professionals in science and engineering.
i Group conducted in-house and contract research into nuclear store generator secondary side corrosien problems, evaluated plant chemistry data, maintained CE's Chemistry Manual, and responded to plant problems.
6/73-6/77: Senior Engineer, Chemistry Development L
Performed theoretical analyses of chemistry / chemical engineering /
corrosion prcblems. Planned and conducted special procedures at operating plants to evaluate system performance and to resolve problems.
i U.S. NAVY, NUCLEAR POWER SCHOOL, BAINBRIDGE, MD.
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4/69-4/73 Naval Officer Instructor of Chemistry, Metallurgy, and Radiation Fundamentals w
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o EDUCATION:
CLEVELAND STATE UNIVERSITY, 9/62 - 6/67 BS Chemical Engineering (Magna cum Leude)
Co-op Experience:
- 5. F. Goodrich Chemical Co.
UNIVERSITT OF MENNE50TA, Fall 1967 UNIVERSITY OF DELAWARE (Evening), 9/71 - 6/73 Graduate studies in Chemical Engineering AFFILIATIONS :
Tau Beta Pi-Honorary Engineering Society American Institute of Chemical Engineers American Nuclear Society REPORTS
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o Member, Steam Generator Owners Group Water Chemistry Gaidelines Commaittee.
Contributor, "PWR Secondary Water Chemistry Guidelines," September 1981.
Beineke, Morgan, Hall, Marugg, Wiatrovski, " Test of Isothermal o
Soaking Procedures for Limiting Tube Denting in Nuclear Steam j
Generators, "EPRI NP-1761, April 1981.
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EPRI Quarterly Reports, RP623-2 Neutralization, RP623-3 Condensate Polishing, 1979-1981.
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" Evaluation of Reactor Coola'nt Crud Samples," CE Internal Report, 3/80.
" Test Report, Hydrogen Absorption (Efficiency in the VCT)," C5 o
Internal Report, 11/76.
"An Evaluation of the Influence of Letdown Degassification on the o
Control of RCS Hydrogen Inventory with Hydrogen Overpressure on the Volume Control Tank," CE Internal Report, April 1976.
o "An Evaluation of Sodium Pass through in Nuclear Once-Through Super-heating Steam Generators Relative to Solids Deposition in Turbine Cycles," CE Internal Report, 12/75.
o "An Evaluation of Core Crud Deposition and its Effect on Core Pressure Drop," CE Internal Report, 4/75.
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" Fort Calhoun Reactor Coolant System Peroxide Treatment, November 1974, "CE Internal Esport, 2/75.
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" Analysis of Core Crud Samples,." Various CE Internal Reports, 1973-1976.
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l Richard W. Weeks Associate Director Materials Science Division Argonne National Laboratory 312-972-4931 Responsible for the materials technology programa at Argonne currently involving the efforts of about 120 people. Materials reliability problems are addressed in a wide variety of energy systems including conventional and advanced nuclear reactors, and coal conversion and combustion plants. Work on stress corrosion cracking in LWR systems has been conducted for more than six years under EPRI, NEC, and direct utility sponsorship. Member of EPRI Corrosion Advisory Committee from 1975 to 1980 and currently a member of the EPRI Materials and Corrosion Committee. Member of the DOE Nuclear Systems Handbook Advisory Committee. Author or co-author of over twenty technical publications and member of the Editorial Board of the Journal of Nuclear Engineering and Design. Registered Professional Engineer in Illinois.
B.S.M.E. Swarthmore College,1964; M.S.M.E. Caltech,1965; Ph.D. Theoretical and Applied Mechanice U of Illinois,1968.
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RESUME 0F ARTURS KALNINS PERSONAL DATA I
Date of Birth:
February 13, 1931 EDUCATION 8.5.
'Eng. Mech.
The Universit'y of Michigan 1955 M.S.
Eng. Mech.
The University of Michigan 1956 Ph.D.
Eng. Mech.
The University of Michigan 1960 EMPLOYMENT C'
1958 - 1960; University of California, Berkeley; Research t
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Engineer 1960 - 1965_'
Yale University; Assistant Professor 1965 - Present Lehigh University; 1965 - 1967, Associate Professor; since 1967 Professor of Mechanics PROFESSIONAL SOCIETIES Member ASME Fellow Acoustical Society of' America Founding Member Academy of Mcchanics EDITORIAL 804RD t
Associate Editor, Journal of the Acoustical Society of America, since 1970 i
L-CONSULTANTSHIPS
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Has served as a consultant on shell analysis to more than 50 organizations.
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MAJOR SCIENTIFIC ACCOMPLISHMENT l.
Has invented (in 1964) a numerical met. hod for solving boundary value problems, governed by N first-order, ordinary differential equations, and applied it to the stress analysis problem of axi-symmetric shells.
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PUBLICATIONS
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One book and 46 articles in recognized journals, l'
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PERSONAL DATA SHEET -
Name: St. ley Austin Holland Social Security Number:
Title:
System Production Engineer Date of Employment:
10/16/48 Education:
j High School - Cool Springs High School Forest City, North Carolina
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ICS Correspondence Schools - Power Plant Engineering Full time curriculum in physics, nuclear physics, I
trigonometry, and chemistry prior to entering reactor L.
operator training - 7 months.
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Full. time participant in the Westinghouse reactor operators training program covering pressurized water reactor technology, nuclear physics radiation control, calculus, chemistry, and PWR design - received j
operator license certificate - 9 months.
B&W pressurized water reactor design, strength of material, basis for design, operating characteris-i ties - 4 weeks.
General bectric Company - steam turbine and generator operation.and design criteria.
Full time training associated with receiving NRC l
reactor operator and senior reactor operators license. Various engineering related studies as-sociated with maintaining license.
l Extensive training in health physics, system opera--
tion, rad waste operation, simulator and control room j
operation.
Numerous supervisory and management development pro-grams sponsored by Duke Power Company and Inter-
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national Management Council of Charlotte, N.C.
Parti-cipating in EIT review course sponsored by UNCC.
l Experience:
1976 t'o Present - Approximately 6 years experience workirig with the General Office nuclear station L
operation support group. Provides day-to-day opera-tion and maintenance support, active involvement in outage planning and execution, vendor contract.
negotiation and support coordination, station inter-face with Design Engineering and other groups within Duke Power Company, interface with other utilities.
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- implementation of a computerized outage management I
program for Duke Power Company nuclear stations, implementation of the B&W ATOG program, working with IMPO to develop an Emergency Operating Procedure Writers Guide - holds Crisis Management Team position.
Responsible for review and implementation of vendor
- recomendations concerned with station operation.
1968-1976 - Various start up responsibilities for all three Oconee units - assigned full responsibility for completing construction, hot functional testing, and start up of Oconee Unit 2.
Held reactor operator and senior reactor operator license. Held positions of shift supervisor, assistant operating engineer, F7 and operating engineer. Assisted in the formation L,
and development of the Oconee operation organization and operating procedures.
j Prior to 1968 - Approximately 20 years fossil station experience in boiler, turbine, and control room operation.
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STEPRE*f D. BROWN i
CURRErr PCSITION I
-l 1980 - Present PRINCIPAL ENGINEER - INSPECTION APPLICATIONS DIVISION - J. A. Jones Applied Resesech Company, EPRI NDE Center, Charlotte," North Carolina. Responsible for Eddy Current and Acoustical Holography technology transfer.
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CAREER EXPERIENCE 1974 - 1980 GROUP LEADER, FABRICATION & QUALITY I
ASSURANCE - Battelle-Columbus Laboratories, Columbus, Ohio 1967 - 1974 RESEARCH ENGINEER, ELECTRONIC j
COUNTER MEASURES GROUP - Rockwell International, Columbus, Ohio i
8 EDUCATIOK The Ohio State University, 1974, Degree Electrical Engineering The Ohio State University,1967, B.S. Physics i
PROFESSICAAL ASSOCIATIONS t.
Chairman of ASME Multifrequency Eddy Current Task Group
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Alternate on ASME Working Group on NDE I
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Stephen D. Brown L
Mr. Brown has 14 years experience specifically related to NDE technology and r-I closely related disciplines. As a Croup I.aader at the Battelle-Columbus
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I.aboratories he uns c'esponsible for the development and implementation of eddy current research programs which included the technology transfer of multifrequency/multiparameter eddy current techniques for the inservice inspection of PWK steam generator tubing; the evaluation and quantification of existing single-frequency eddy current technology for PWR steam generator
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tubing inspectien; design and construction of fixed-site and air-transportable PWR steam generator mockups for the training of inservice inspection personnel and evaluation of existing and potential inspection techniques. Mr. Brown has also been instrumental in the application of com-puter modeling techniques for the optimization of planar multilayered eddy current problems and the selection of optimal mixing frequencies for the multiparanter eddy current inspection of tubing.
i Earlier at Battelle, Mr. Brown accrued extensive experience in the boresonic inspection of steam turbine rotors and is co-author of a report documenting
[
boresonic state-of-the-art capability. He uns also a member of the Battelle rotor analysis group responsib~4 for the third party assessment of rotor remaining lifetime and has analyzed approximately twenty-five rotors to-date.
Mr. Brown was also involved in the early assesament of BWR piping ultrasonics inspection methods and has extensive ex,:erience in the develop-ment of ultrasonic inspection techniques and systems for the inspection of deplaced uranium penetrators.
Mr. Brown has also conducted independent research activities in the area of coherent optics and has implemented L
speckle diffraction interferometry tachniques for the measurement of in-plane displacements and strain and time-average holography for the measure-sent of resonant modes of vibrating structures.
At Rockwell International, Mr. Browtf was a microwave systems engineer responsible for the analysis of foreign radar-associated weapon systems, the e
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assessment of U.S. derived electronic countermeasures, and the development of advanced countermeasure systems.
t i-L e.
-.e5
E',
W. H. Layman is at present the Department Manager for Generic Safety Analysis in the Nuclear Safety Analysis Center (NSAC) operated by the Electric Power Research Institute (EPRI).
He has been in the nuclear power industry since 1952.
During his nine years in the nuclear submarine program Bill served as Chief Operator in starting up the first submarine proto-type nuclear power plant in Idaho, then served as Assistant Engineer Officer in the initial crew of the first nuclear submarine, Engineer Officer and Executive Officer of 12.ter I-.
nuclear submarines and Squadron Engineering and Material Officer for the first squadron of missile firing submarines.
f' l
. In 1961 Bill joined the Pennsylvania Electric Company where
'he served as General Manager of the Saxton Nuclear Experimental Corporation and then as Generation Division Manager of Pennsylvania Electric Co.
In 1960 Bill joined the Atomic Energy Commission as Chief, j
Water Reactor Branch, Division of Reactor Development Technology.
In 1973 he was promoted to the position of i
Assistant Director for Operations, Division of Reactor Safety Research.
From 1975 to the present time Bill has been a member of the EPRI organization where he has served in a number of capacities including Director, Steam Generator Project Office; Associate Director of NSAC and Department Manager of the Plant Engineering Department.
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APPENDIX C
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b TRIED PARTY REVIEW OF THI-1 OrSG IEPAIR PROGRAM CIAITER I.
PURPOSE i
It is the intended purpose of this " Third Party Review" (TPR) to provide a timely, independent, objective, safety evaluation of all activities defined in this charter for conformance to:
- 1) the NRC rules &
regulations governing the operation of IMI-1; and 2) the adequacy of the steam generator repair program that will allow safe operation of the nuclear unit.
II.
SCOPE The scope of this review is generally limited to activities associated g
with the identification.of failure mechanisms and repairs of the TMI-1 Once Through S team Generators (OTSG's).
The specific task areas to be reviewed are described in more detail in Section IV of this charter.
It is t'.no intent of GPUN Management to fully develop and implement repairs to the 1MI-1075G's within the provisions of 10CTR50.59.
It is expected that the TPR will promptly notify GPUN Management of any circumstance not already identified by GPUN, that fails to meet these standards.
III.
MEMBERSHIP The membership of the TPR body shall include individuals with expertise in the following specialty areas:
A.
Steam Generator Design and Performance
- 1 member E.
Chemis try
- 1 member C.'
Materials
- 1 member
~
D.
Stress Analysis
- 1 member E.
Safety Analysis
- 1 member F.
Plant Operations
- 1 member G.
- 1 member The TPR shall have a GPUN individual assigned as Secretary for the review.
He will be the general interf ace for the TPR membership and GPUN. A Chairman will be elected and report the results of the revicv
~"
directly to the Vice President - Technical Functions in this assignment.
The Secretary will arrange for all review material, meetings and recordkeeping for the team.
Any specific information requests within the scope of the TPE should be to the Secretary. The Secretary will be a non-voting member in any matters of the TPR seeking concensus opinions.
say. 3, 4/27/82 g
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~
Members of the TPE will cither be from outside the CPUN organization or from the portion of GPUN not responsible for the steam generator repair or 'D11-1 operations.
IV.
SPECITIC REVIEli AREAS A.
Failure Analysis Program I
This program is intended to identify the cause of tube cracking and means to arrest it.
The program will also include an evaluation of other portions of the reactor coolant system to determine if corrosion mechanism extended out of the steam generator boundaries.
3.
Eddy Current Examination Program This progran is to develop ami imp'lement an eddy current. examination
'~
method to idsntify the extent of the tube cracking problem.
C.
OTSG Perfo mance Evaluation I'
It is the object of this effort to evaluate the impact of the repair
~
procedures on the performance of the stema generators, especially in the area of safety analysis.
D.-
Repair Criteria This program is intended to provide guidance concerning ttie~^tyle7
-~
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repair to be done on the damaged tubes.
E.
CISC Repair Program This program covers the actual repair of the steen generators.
V.
MEETING FREQUENCY i
The TPE shall initially meet to receive a presentation by GPUN and its consultants on the current status and direction of the OTSG Repair i
Pressam.
At that meeting, they will be presented with initial reference material that will enable them to assess product.s then available.
i A second meeting will be scheduled by GPUN to make a final presentation to the TPE prior to initiation of the final productiou repairs.
Other meetings may be held at the discretion of the TPR.
YI.
_RECotDS The Secretary shall maintain records of all meetings of the 'TPL The Secretary shall also maintain a record of all documents reviewed by the membership.
Portions of the material may be proprietary in nature.
i Appropriate arrangements shall be made to protect proprietary information when it is used.
Rev.'3,4/27/82
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Transcripts may be taken of the final review meeting and used as part of the documentation package available to the NRC in support of the OTSG Repair Program.
A final report will be prepared by the review team which sunanarizes its findings and conclusions regarding the safety adequacy of the repair program.
The report should a.ake an explicit finding that the approach proposed by GPUN is adequate if that is the conclusion of the review.
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1 APPENDIX D j
l.-
I TPR Bibilography 4
W e
1.
Date: April 12, 1982 r
Subj ect: Third Party Review To: TPR Members o
l ;--
Material Transmitted:
t'
.l' OTSG CHARTER 4
it
.2 Membership List and Resumes of TPE
- j' J Map to GPUN 4 OISG Repair Task Organization J Preliminary Report Failure Antlysis r"
4 Preliminary Report - Eddy Current Exam Program 4
{
7 Agenda - April 23 - TMI-l OTSG Review Outline J Handout, from April 7 Status meeting with NRC 9 Handout from January 25, TMI-l O'SG Status Review (NRC) 2.
Date: April 16,1982 o
Subject:
TPR To: TPR Members Material Transmitted:
1I
.1, Part I, Handouts from April 7th Status Meeting 1 ' '-
with NRC*
.2 Two Additional Resumes of Members of TPR Taam
{( l--
e 3.
Date: April 23,1982 (No Cover Lette'r)
Subject:
'TPR l,
To: TPR Members L.
~
Material Transmitted:
i i [~
1 TMI-1 Steam Generator Recevery Program Task 7
.2 Essetor Coolant System Inspection and Requalification
- j (April 16, 1982) B&W i
- l-
-3 Bandout from NRC April 23, 1982 from N. Kazanas
.A Reactor Coolant System Review Task 7 4/23/82 1'
J TMI-l CISG Tube Failure Probability from D. Slear 4/23 1
6 OTSG-B E/C Absolute (4 x 1) Results Suasaary from D. Slear 4/23
'L
.7 TMI-l CISG Tube Making Procese 4/23 j
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e 4.
Date: May 13, 1982 i
Subject:
Draft Minutes of First Meeting
~.-
To: TPE Members t
}
Material Transmitted:
.1 OTSG Charter (Revised)
~
.2 Preliminary Specificatic,n for the Repair
.3 THI-1 Tube Preliminary Stress Report (4/21/82)
-F 4 MPR Associates - draft Residual Stress Report (April 27, 1982)
J RPRI Reference Material I r.
a.
RPRI Activities,in Support of THI-l CISG Recovery l
Memorandum Report April 1982
,L
'b.
APPENDII 1 Status Report 2/24/82
}_
c.
APPENDII VI Laboratory Experience of Cracking of Materials (Other than IN600) in Solutions Containing Sulfur Species e
3.
Date: June 1, 1982
~
Subj ect: Draft Minutes of May 20, 21 TPR Meeting To: TPE Members Material Transmitted:
.1 Telecon M. Graham w/R. Jacobs (NRC) 5/24 re: NRC Participation
.2 Memo: DISG Testing Program 4/1/82 TMI-E3914 from D. Slear
?
e 6.
Date: June 21, 1982
Subject:
TPR To: Prof. Arturs Kalnins l
1 Material Transmitted:
I"
.1 EPRI Letter: Status of Taskt Assigned to RPRI by GPUN Failure j
Analysis Task Group on 2/11/82 e
7.
Date: July 2, 1982 3
Subj ect: TPR l
To: Prof. Arturs Kalnins
,) _
Material Transmitted:
.1 Stress report Work by Jim Moore (Updated Copy)
?
3 l-e 8.
Date: July 26, 1982
Subject:
TPE j, f,,
To: Prof. Arturs Rainins
,u Material Transmitted:
r i
.1 RAW - 10146 (October,1980), " Determination of Minimum i
~
Required Tube Wall Thickness for 177 FA Once Through Steam Generators" t
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9.
Date: August 10, 1982 j
Subject:
TPR To: TPR Members d
Material Transmitted:
3
!'l
,1 1MI-1 OTSG Failure Analysis Report, 7/82 I
2 TMI-1 OfSG As-Built Tube Stress Analysis
.3 1MI-1 OTSG Recovery 4 34W Evaluat' ion of Tube Samples from TMI-1
.5 Battelle final report on Fail.are Analysis of Inconel 600 Tubes from OTSG A and 5 of TMI-1
, jr l
e 10. Date: August 20, 1982
' Subj ect : 1MI-1 OTSG Repair Safety Evaluation I
To: TPR Members
- s. t__
g Material Transmitted i
.1 TMI-1 OTSG Repair Safety Evaluation
~
e 11.
Date: September, 1982 Subj ect: Minutes of Meeting fo: TPR Members Material Transmitted
.y
.1 Sulfur Cleanup Handout
.2 OISC Logic Diagram i!
.3 B&W Memo of July 30, 1982 from J. F. Pearson:
Report on
,8 Prequalification Charge Sizing for TMI-1 Steam Generator I,
Tube Expansion
.4 Eazanas Randout: August 9, 1982 THI-1 OfSG Task 4 Eddy Current
- I Presentation to the NRC.
1...
A e 12.
Date:
September 17, 1982
Subject:
THI-1 OTSG Repair Process Description and Qualification l
To: TPR/GORB Members
'l j!
Material Transmitted:
1 Randout from September 15 meeting (in conjunction with
- i Foster Wheeler and Babcock & Wilcox) with NRC.
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e 13. Date:
September 24, 1982 I [-
Subject:
TPR To: TPR Members t
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Material Transmitted:
.1 Revision 4 of OTSG Tube Repair Specification i
e.14.
D te: October 5, 1982
Subject:
TPR To:
W. R. Layman (cc: TPR) t Material Transmitted:
i I' ^
.1 Transparancies used during TPR Meeting held on 8/25/82.
E
.2 Susanary text used as mechanical input to SER.
fr-J Prof. F. Erdogan's (Lehigh University), Part I of II.
- 15. 'Date: October 20, 1982
~~
Subj ect: IFR To: TPR Members Material Transmitted:
u-
) l..
1 Revision 1 to the Steam Generator Repair SER.
.2 NRC's SEE on the Explosive Espair Process.
},{ 7-3 (Via D. G. Slear) Presentation material given to NRC during 2
10/16-10/19 meeting.
e 16.
Date: November 19, 1982 Subj ect: TPR To: TPR Members 2
l.. f-Material Transmitted:
',Ll -
.1 Draft Safety Evaluation for Return to Service after OTSG 7-Repair
.2 Results of Argonne National Laboratory's analysis of the l g' Sludge in the "C" Reactor Coolant Bleed Tank" 3 Draft Guidelines for proceduras to deal with beyond design basis tube ruptures.
l..
e 17.
Date: December 1, 1982 T'
Subj ect: TPE L
To: TPR Me=bers l,
Material Transmitted:
t.L_
1 THI-1 OTSG Tubing Eddy Current Program Draft Qualification Report (October, 1982)
. I,i 2 GPU Nuclear Memo: Procedures Plan for Issuance of I I2 Guidelines for Single and Multiple Tube Ruptures, 1
Dated: October 13, 1982.
, l' 3 GPU Nuclear
- Memo: Progress Report on OTSG Tube Rupture
! L_
Procedure Development, Dated: November 19, 1982.
I b_,
O m.
amm..
i
.4 MW Final Draf t Report; 'DfI-1 OISG Repair Kinetic 4
Expansion Technical Report, Rev. O, Deted: November, 1982.
5 Response to TPE Questions.
.6 Results of Supplemental ISI.
7 B&W, Research & Development Div Letter, Emport No. 5433-01, July 13,1982.
.8 TF1 Bibliography.
9 Cafety Evaluation for Return to Service Af ter Steam Generator Repair, Rev. O.
.10 SDD 232-C, Rev.0, 'DtI-1 Erimary to Secondary Leak Temporary Waste Processing System.
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APPENDIX E
'COM ILATION OF REVIEW GROUP OUESTIONS & GPU NUCLEAR ANSWERS
.I.
PLANT OPERATIONS i.
1.
Are there plans to augment ~ operating shifts with additions 1 people j'
specifically for responding to steam generator tube leak incidents?
There is no need to assign additional personnel for responding to steam
,I generator tube leak incidents. This incident can be handled by normal staff on duty.
A tube leak incident of greater than 1 gym but less than 50 gym r
would be categorized as an alere by the unit Emergency Plan.
A rupture of greater than 50 spa is a site emergency.
Both levels of event require implementation of the emergency plant and organization.
The supplemental 4
j-staff will be available on site within 60 minutes, i
i"_
2.
Ia' additional training required for operating shift and support personnel to better prepare them for managing tube leak incidents?
I Yes. The following training has been conducted or is' planned to improve handling of steam generator tube leaks:
,l a.
Following the Ginna steam generator tube rupture, all operators were trained in accordance with NRC guidelines and using the lesson program developed at Rochester Gas & Electric.
'{l b.
The annual simulator requalification training for operators is scheduled for January and February, 1983. As was done last year, both simulator and classroom time have been set aside for steam generator tube leak incidents.
.I This year additional classroom hours are scheduled to discuss the upgraded steam generator tube rupture procedures.
1 Training is planned at TMI on the upgraded procedures c.
as they are 5*~
finalized.
4 j
Training will be provided on operation of all new equipment.
d.
h 3.
Are there presently 'any indicat ins that failed fuel may be a problem? Is i
there a procedure that will identify aC system activity increases during low power tasting?
At the end of the last cycle we had 0.03% failed fuel compared with 1.0%
j, f ailed fuel used in accident analyses. This was the higest experienced at TMI-1.
With a beginning-of-cycle core now in place, the percent failed fuel should be substantially lower.
RCS activity is routinely monitored and an l>
increase in activity would be evident during low power testing.
,'[l 4.
Station capability for b.ndling leakage from the secondary system in event of a significant steam generator tube leak experience.
? l_
?
- a. Installed and proposed processing equipment to handle contaminated water input to the turbine building sumps (valve or other leakage from the secondary system).
Additional processing equipmeat and a holdup tank are being installed to l
provide sufficient capacity and operational flexibility in case of I.
- 1.. __ _ _...._..
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,j significant S. C. tube leak. This
- modification includes installing pipes from the TMI-1 Turbine Building sump to a 250,000 gallon tank.
?
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The system and its intended operation are described in detail in Ref.17.10.
?
?.
- b. Capability to reduce, eliminate, or re-direct cooling water or other inputs to the turbine building sumps during shutdown and cooldown of the unit to minimize the volume of water to be processed.
The following drains have been or will be rerouted prior to return to service:
' l p-1.
Industrial Cooler Blowdown Drain.
l.
2., Main Stata Safety Valve Drain.
j -'
3.
Aniliary Boiler Blowdown Sump Pump Discharge Drain.
{
4.
Secondary Plant Sample Sink Drain.
- l There modifications reduce the volume of water collected in the turbine building sump.
In the event of a large steam generator tube rupture, the
{'
fluid that would normally be processed through the turbine building sump
[,
will be pumped to a large hold up tank. The new tankage will hold up to 30 days fluid for processing.
U c.
A r.e there installed systems that will minimize contaminatisn of the auxiliary steam system med boiler?
3i No.
Aux boiler uses condensate.
Leakage from aux boiler to its sump is to i.
be directed to Turbine Building sump for monitoring prior to discharge and i I processing, if required.
J '.
j[
d.
Capability for inservice cleanup of secondary system contaminated water
,y prior to restart of the unit.
- F-
- .g Yes.
This capability will be provided using Powdex Resin System and g-disposal of Powdex resin as radweste. See additional information in answer to I.4.a.
6
- L.
e.
Approved procedures and operator training required to minimize off site releases from secondary system contamination.
Procedure guidelines which address normal offsite releases and offsite j~
releases due to tube ruptures are planned for completion prior to 12/31/82.
Training for the procedures is planned to begin prior to 1/15/82.
L
?
5.
Will procedures dealing with multiple tube failures be available?
,1
.l Yes.
See draft proced:2re guidelines for beyond DBA procedure development plans (Esf. 16.3, 17.2 and 17.3).
u
~
6.
Do you plan to address simuli.aneous leaks in both generators?
Ii" Yes. See draft procedure guidelines (Ref. 16.3,'17.2 ana 17.3).
~
7.
In event of a tube rupture and auto initiation of high pressure injection l ;;
i
~
~
1
- ~ - - -
I
normally all E pumps will be removed from service.
Will natural circulation provide sufficient cooling to prevent steam reliefs from opening and what is the impact of this steam release?
Response to this ites will be supplied at the December 7th meeting.
8.
Is it possible to completely cooldown and depressurize the primary system, without reactor building entry, to minimize primary leakage into the secondary systes due to differential pressure?
r
'[
The equipment usad to minimize RCS/0TSG differential pressure is dependent on the RC pump status. With BC pumps available pressurizer spray would be ip used to maintain the minimum allowable subcooling margin (i.e., minimize OTSG 1eakage).
L With RC pumps unavailable, then RCS pressure would be controlled by either the pressurizer vent line or the PORY. Both of these valves are remotely l-operated from control room.
Therefore, containment entry would not be required to minimize OTSG 1eakage.
,4r
-l 9.
The OTSG Task Organization included in Task 8 an evaluation of modifications that may be necessitated by operation with small primary to secondary leaks.
However, the agenda for the review appears not to include discussion of the modifications that asy be required.
The three categories' of planned modifications are as follows:
i i
TMI-1 OISG Task 8 Modifications 9
A.
. Radiation & Contamination Control Modifications
[~
Temporary shielding (as required), lining of sumps, reroute of some drains,
i.-
airborne & liquid leak detection / monitoring upgrade of industrial waste
,I treatment system and filter system (IWTS/IWFS).
t.
B.
POWDEI Resin Processing / Backwash Water Recovery Modificat' ions Portable skid mounted.
L'-
C.
Increased Water Storage Modification i~
Holding tank for abnormal / casualty conditions ( 250,000 gallons) with connections for portable domineralizers.
These three modifications are described in detail in Ref.17.10.
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I I c-f II.
SAFETY ANALYSIS f
l 1.
Reactor Control and Shutdown j
a.
Has in-core instrumentation been affected by the corrosion process? (in a
- way which might invalidate safety setpoints, or prevent a safe shutdown).
i No.
It has not been affected. The Task 7 final report CRef. 3.1) reports the results. See Table V-1 sections:
5' 1
i VI.,
i.e., VI.2.b, VI, 2.d, VI.3.e l ; ;' -
b.
Have control rod drive mechanisms been affected so that they will not i
I perform adequately their control and shutdown functions?
'f l
The Task 7 final report (Ref.3.1) reports the results.
See Table V-1' sections:
VI.l.b, VI.1.c, VI.2.a.
I' I
c.
Has corrosion increased the probability of a control rod ejection accident?
The control rod drive mechanism is mounted on an inconel nozzle velded to the carbon steel reactor vessel head.
Stainless steel clad is buttered over the bi-metallic weld.
We have concluded that this is unlikely based on our inspection program.
l{
We have found no damage in that area.
s See inspection of CIDM In Task 7 final report (Ref. 3.1 Section V).
l d.
Could the corrosion process have degraded the ability to shut down the reactor by baron injection?
j Charging system (LPI/HPI) nozzles were inspected.
No damage was found.
These areas represent conditions of highest susceptability.
2.
Reactor Coolant Systes Inventory
,,i !.
a.
Has the primary coolant boundary been degraded in areas in addition to the l-steam generator tubing? Examples of susceptible areas include reactor vessel
[
nozzles, safety valves, PORV, reactor cociant pump seals, and isolation valves.
Nothing has been i.:entified. See Task 7 report (Ref.3.1 i
j Sec'. ion V) for details.
b.
Has the ability to make up fluid inventory been affectai?
(tanks, piping, j
valves, pumps) u In addition to the Task 7 inspection program, supplemental ISI has been
-performed on Spent Fuel, Decay Heat and Building Spray systems.
No degradation that would lead to an inability to make up fluid inventory has
~
i been found (Ref. 17.6).
3.
Beyond 7esign Basis Tube Rupture l
i..
l _.
a.-
1 a.
Should safety analyses be performed assuming multiple steam generator tube ruptures? The normal design basis analysis considers the guillotine rupture of j
only A single tube. The basis for performing analyses of only a single tube rupture has been that it would be caused by a random failure and that operation would not be allowed with the general condition of the tube bundle degraded.
With slower developing f ailures it has been demonstrated that inspections and l
tight leakage limits allowed degraded tubes to be plugged before serious failure and before a large number of tubes could degrade to the point where multiple
- ; g-tube failures would be considered to be credible.
With'.the potential for l6 l.
11cipient cracks in a large number of tubes we should evaluate the probability l 1' taat the plant operators might be faced with :ultiple tube f ailure in a single accident.
It is probable that simple analyses will show that the consequences of multiple tube failures (dozens or more) within the tubesheet area are bounded I
by the design basis single tube rupture out of the tubesheet area. This result could give us high confidence that even if the proposed repair were to be faulty, the great majority of cracked tubes which are within the tubesheet area would not pose an unacceptable safety probler.. Incipient cracks below the upper tubesheet pose a more difficult problem.
Based on our Safety Evaluation (Ref.17.9,Section II), we have concluded that ECT has detected those cracks with the potential for propagation to failure during a transient or due to flow-induced vibration.
We conclude that an7 failure which might occur in the future could be characterized as
.randon.
In addition, leakage through cracks of various circumferential extent has b e en analyzed to determine a thr.eshold of detectability.
Analysis shows that a crack wuld be identifiable due to leakage before it
[
propagated to a size that would rupture under accident loading. (See also Ref. 17.9, Chapter II).
t-
,i However, we are developing a program for multiple tube ruptures and ruptures j"
in both generators. See draft guidelines (Ref. 16.3, 17.2 and 17.3).
![
l b.
Are existing design basis safety anaIyses affected by this corrosion j {' i problem? Examples could be main steam line break analyses and loss of coolant accident analyses.
A pertinent question is about the possibility of cas of these design basis accidents being omplicated by having the force of that accident cause simultaneous tube f ailures in a degraded steam generator tube bundle.
l The design basis analyses have been applied to both the new joint and the tube portion left in. service.
The new joint has been qualified to hold under maximum accident loading (main steam line break) with margin. Thus
,*j aultiple tube ruptures in the tubesheet caused by joint loosening under
- L.
accident condition are considered unlikely (See Ref.17.9 Chapter V and Ref.
17.4).- Analyses have also been done to determine the size of crack which l, 'I would rupture under accident loading.
u Analyses of cracks caused by this corrosion problem indicate that a propagating crack would be identifiable due to leakage before reaching a size that would ductilely fail during a main steam line break.
(See Return
~
to Service SR1 (Ref. 17.9) Chapter II for further details.)
c.
There are a number of hypothetical "what if" questions that need to be 7
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ijj addressed. For example:
3 How would IMI-1 respond if (a) several tubes burst essentially simultaneously in
,M' either of its steam generators, or (b) tube failures occurred in both of its
]j steam generators simultaneously? I realize these questions may be outside the
- ?
current design basis, but in view of the extensive tube damage in both steam generat' rs and the novel "fix" proposed, these questions should at least be o
discussed.
l..
A program addressing TMI-1 response to multiple tube ruptures and ruptures in both generators is under development. See draft guidelines (Ref. 16.3, 17.2 and 17.3).
2.
i 4.
Has the corrosion process degraded the ability of the plant to remove heat f,j i.'
from the reactor and transport it to heat sinks? This includes the ability to l y, shutdown from operation, cool down, and maintain the system in a cold shutdown 2;
condition (heat exchangers, pumps, piping, and valves.)
i Task 7 examined a sampling of areas of expected highest susceptability.in tha primary system.
See Ref. 3.1, Chapter III, Table III-1 for a listing.
g..
i In addition, supplemental ISI was performed on several supporting sys tems.
(Ref. 17.6).
No damage was found.
Based on these results, it was projected i
1 heat from the reactor.
that no damage has occurred to degrade the ability of the plant to remove j
lj 5.
Has the corrosion ~ process degraded the ability to isolate the reactor j*;
containment building?
Examples might be fluid system containment isolation j
valves.
I:!$;
Based on Task 7 (Ref. 3.1 Chapter Y) results, no damage is projected.
i i!
6.
Is sodium thiosulfate going to be eliminated' from the site?
Since the
{j thiosulfate was present for the purpose of post accident radioactive iodine control, have analyses confirmed the adequacy of iodine control without the use of sodium thiosulfate?
Yes, rhiosulfate has been eliminated from the unit.
The required changes (Tech. Spec. Change Request and associated Safety Analysis) justifying the adequacy of spray iodine retention, have been submitted to NRC. Resolution is required prior to restart.
I
- I L-7.
Radioactivity Releases (a) Ras any equipment been.affected that is used to mitigate accidents?
- L In the Task 7 sampling of equipment (Ref. 3.1 Chapter Y); no damage was i~
found in areas of the primary system of suspected highest susceptability.
L Based on these results, no damage to any equipment has been projected.
(b)
Has the corrosion process caused any condition which could increa'se significantly the release of radioactivity?
Based on our saf ety evaluation (Ref.17.9), we do not anticipate an increased probability of events leading to significant releases of
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.i radioactivity. The repair process itself will leave the steam generators with mechanical rather than' welded joints which can be ' expected to result in a slight increase in primary to secondary leakage.
Thus routine offsite releases may increase slightly, but should remain within Appendix I limits.
1 I
(c)
Will the plant condition have any effect on emergency guidelines or procedures?
Possible effects on emergency guidelines or procedures are being evaluated as part of the overall program to upgrade tube rupture procedures.
See Guidelines (Ref. 16.3, 17.2 and 17.';).
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.It III. NDE.
- 1. GPU believes that if the generator tubing is cracked then the crack is through-wall.
Physically, I find this hard to believe. The Battle 11e i
metallography shows intergranular penetrations 10%-50% through wall. This would seem to demonstrate that we in fact can have a continuum of depths.
1 E I.
Although ECT probes use'd have been qualified to ' find much smaller intergranular cracks (Ref.17.1), the vast majority of cracks identified are
?
through-wall.
However, the existence of unidentified cracks below the
'l threshhold ECT detectability has been assumed in the safety analysis.
See SRR (Ref.17.9) Chapter II.
- 2. What is the largest crack (length-depth) that can be allowed to remain la a tube?
Curves have been developed for length vs. depth, indicating the largest crack that can remain in service.
See SEE (Ref. 17.9) Chapter II.
e 3.
What is the reliability of the existing eddy current NDE methods (differential coil or 4x1) in detecting the crack mentioned in 2)?
?
Both the.540 differential core and the 8 x 1 absolute probe have been demonstrated to have adequate reliabiliy for detecting the cracks described above. See SER (Ref.17.9) Chapter IX.
4.
The lane region in OTSC's is subject to cros s-flow conditions.
Shallow intergranular penetrations may act as seed cracks which may rapidly propagate to failure in a high-cycle fatigue mode.
Should special consideration be given with regards to tube stabilization.?
Tubes to be plugged will also be stabilized if they have cracks in areas of high cross flow.
See SER (Ref.17.9) Chapter II for a detailed discussion of areas to be stabilized.
5.
Provide the basis for the sensitivities used during the in-generator eddy current tests.
See the draft ECT qualification report (Ref. 17.1).
6.
Similar chemical conditious appear to cause both IGA and IGSCC in sensitized i
I-600, and IGA has been observed in cracked areas. Are techniques available to periodically monitor for increases in the extent of IGA as an indicator of i
l potentially corrosive conditions?
l No, not with nondestructive techniques.
Leakage data in operation and ECT
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will detect major changes in tube integrity.
t 7.
Are techniques available to determine the effect of explosive expansion on existing cracks in the tubing?
The effects of explosive expansion on existing cracks are being evaluated 3-
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outside the OTSGs. Tubes removed from the generators and sulfur induced IGSCC grown in the lab have been expanded and examined ECT dye penetrant l
and, destructive techniques.
No ductile. tearing of cracks has been identified.
See answer to question V.1.C.
8.
Are techniques available to periodically monitor for tubesheet corrosion l4 under, cracked areas of tubing repaired by explosive expansion?
1 2 No, not with nondestructive techniques. See answer to question V.1.C.
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'I IV.
Failure Analysis Program (cause & cura) 1.
Update of Failure Scenario Esa the scenario been defined more completely since our last meeting? Have the important perameters been identified?
Eas the scenario been verified by duplication in laboratory testing?
i The final Failure Analysis Report is Ref. 9.1.
Cracking of stressed-sensitized I-600 in boric acid solutioca with sodium thiosulfate concen: rations of 1-10 ppe 52 0 3 at 130*F has been duplicated in the laboratory. The actual TMI tubing was found to be exceptionally susceptible ll ;-
to this type of attack when compared to archive samples of the same heat.
(See B&W, Research & Development Div. letter Report Number 5433-01 dated July 13,1982 (Ref.17.7) and Failure Analysis report (Raf. 9.1) for further
,~
details.)
2.
Update on Proposed Sulfur Removal Procedures I
Peroxide creatment has been proposed as a means of removing residual sulfur from i
RCS system surfaces. Given what is known about the failure mechanism, I think this concept needs critical review. My questions and their implications are as follows.
a) What is the basis for believing that sulfur removal is necessary?
Sulfur exis ts on the s team generator tube surf a~ces in rath'er large 2
concentrations (a.3000 pg/ft ).
Analysis of the tube surfaces has shown the sulfur in reduced Jeates as well as fully oxidized. Sulfur.4 also present on other plant surfaces although at lower concentrations.
The oxidation state of the other sulfur is unknown.
If during plant startup the reduced sulfur is slowly oxidized it may pass through a very corrosive stage where additional corrosion damage may occur.
If the sulfur is oxidized, solubilized and removed by the perioxide creatment under protective high pH conditions the potential for corrosion damage is less than if nothing is
'done.
b) What is the basis for believing that peroxide can oxidize sulfur residue all the way to sulfate without pausing at an incarmediate, possible corrosive stats?
The sulfur undoubtedly does pass through the corrosive intermediate state during the peroxide treatment. The breaker experiments conducted at 34telle i
indicated that this transition is rapid.with peroxide present so that r.o new attack is initiated.
In addition, the pH will be slightly alkaline (8.0-8.5) in an effort to further reduce corrosion potential. Corrosion tests are being run to demonstrate the capability of the process to saf ely r
l' remove the sulfur without adverse effects.
t c) What are the application conditions (concentration, pH, other chemicals, time, temperature, flow) and what are they based upon? Are they consistent with the cuidation kinetics?
Tests are being' carried out to identify the optimum application conditions.
L Based on taats with actual TMI-l tubing NiS; the time period appears to be
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s 200-400 hrs; the temperature 130*F; the pH adjusted to. 8.0-8.5 with NH ; the i
peroxide concentration at about 20 ppa. 130*F is the temperature of ten used i e-in operating plants for peroxide addition. Higher temperatures might cause 3
concern for plant safety while lower temperatures will reduce reaction j'
races.
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d)
How is residual peroxide to be removed? If by hydrazine, what is the risk i
I of reducing non-removed sulfur to a possibly corrosive state?
i Residual peroxide will decompose to oxygen which will be reduced by desassing in the normal manner prior to startup. Hydrazine additions sill be made in small increments to remove residual oxygen to avoid establishing p' L a strongly reducing atmosphere.
It may be possible to maintain pH high until deoxygenation has been at least partially complet.ed.
u.
a, e) How is the end point to be determined?
5 End point determination will be based on time and growth of SO4 concentration in the coolant while cleaning.
When SO 4 analyses are i
essentially constant, the cleaning process is considered complete.
i f) What steps will be taken to preclude chloride SCC of stainless component l-while peroxide is present?
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Chloride will be kept at40.5 ppa during the cleaning by operation of the normal purification system. It will be reduced to<.0.1 ppa prior to heatup.
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g) What inspections will be performed following this treatment to check that
- .r further corrosion has not occurred?
)1.
j Further corrosion is expected to be identifiable by increases in leakage.
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My basic problem is that the failure scenario identifies oxidation and/or
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reduction steps as actors in the corrosion. process.
Peroxide treatment may replicate these processes.
If the residual sulfur is present as sulfide how do we know that perioxide treatment will not convert it to corrosive sulfur species on the way to forming water soluble sulfaces?
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See previous response to question IV, 2.b sad c.
i 3.
It would be of use to hear an update on any new information regarding the organic residue in the bleed tank.
u.
b.
See results of Argonne National Laboratory anclysis (Ref.16.2).
4.
How are the concerns regarding sulfur species remaining in the primary circuit (Ref. CPU Appendiz III) to be resolved so that the problem is not
.re-initiated during service? 0xidation with H 2 02 was suggested.
Has an experimental program been initiated to investigate this approach along with any possible side effects? What are the levels of sulfur species in the primary system (on metal surfaces and in the water) at this time? Would it be possible to monitor the electrochemical potential of inconel and stainless steel in the L
primary system water as a function of time during requalification testing?
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to investigate the removal of The cleaning procedure has been des An. extensive investigation has been undertaken using an H2 02 process.
Results indicate that the plant coolant conditions will not i ide at shutdown.
200-400 hours.
gulfur many Westinghouse plants that add per oxcomplished in aboutand af ter kine that complace sulfor removal can be acalso conducted with Immun bs expension, to evaluate the impact of these var l aning process. It does Tests were However, the The Immuno 1 has no appreciable effect on the c e unexpected benefit.
d residue.does slow down the sulfur i
to retard the peroxide decomposit on, an explosive expansion process or uncleane I. _
400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br />.
removal process from about 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> to cess (304 SS-sensitized and de from TMI-1 cubes) have Corrosion specimens expanded to the cleaning p f
shown no evidence of corrosive attack.
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/82 have been analyzed for measurement.
\\ amples of water taken from the RCS on 9/25 Thiosulfate was u lfate S
thiosulfate and sulfate.
I was 30 ppb.
contaminate levels.
i Swipes of plant surfaces have shown the follow ng 2
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522 micrograms SO /ft 4
Fuel Rod 418 micrograms SO /fC 4
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Grid 530-700 micrograms SO4/ft Regenerative Neutron Source Retainer 144 micrograms SO /ft' 4
RNS Spring lfate contamination than new h w some level of contamination,
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These levels in some cases are even lower in suow jL steam generator tubing but in ot,her cases s o l3' although it is still below that chemical potential devices in work was done in a BWR in j,
Very little work has been done using electro i l sidestream loops must be Some operating plants at operating condicons.
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an extensive investigations of pipe cracking btte spec a In addition, eaning of potential variations ilt.
constructed and specialized probes bu j l~
research program is needed to determine t e m the work is done in h
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with various plant conditions changes be oreThis ars for the BWR work.
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i extent of this effort makes it impracticawas part of the c (..f:
its previous function?
e The sodium thiosulfate system f
Now that it has been eliminated, what will per orm 5
Id.
See response to Section II Question II.6. ifies several pos r
ifications The scenario of the failure mechanism identWere considerations
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Je ecies la the future?
6.reduced sulfur species.
t or procedures to preclude presence of these sp include (1) clearer labeling
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New administrative controls which are in ef ect jj
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,h of tanks in the Chemical Addition Room, (2) locking open the breakers to pumps CA-P-2, 3 & 4 and placing them under the administrative control of the 4._
Locked Valve and Component List and (3) review of applicable procedures to insure that adequate guidance is provided. Assuming the effectiveness of lI administrative controls, the only chemical which has a potential for
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inadvertant induction into the RCS is Sodium Hydroxide and this appears l
possible only when the RCS is depressurized.
Under these conditions, F
si additions potentially would not reach OTSG tubing and even in the event that very dilute caustics did reach the tubes, damages would not be expected since the increase in pH value is toward a more benign condition.
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7.
The failure scenario concludes that the time of attack was af ter hot l.,
functional testing in September 1981. No direct evidence is cited to establish
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that corrosion cracking of the tubes cccurred then.
Were there prior I:2 surveillance tests that establish the cracking did not exist at some earlier time?
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lj In 1980 a 3% ECT was performad without evidence of a problem.
Indirect evidence that the Unit was able to function during pressure.
l t-l 8.
Did the hot functional t;ests include any operations that could impose high thermal stresses on the CISGs, such as boil dry and refill or steam generator relief valve testing?
No.
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The Failure Analysis Report indicates thac OTSG performance. history was i.
reviewed in detail.
'It does not state the worst cases found.
Did~ instances, e
such as in 6. above occurf
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. i The review indicated oc severe transients such as in Ques tion 6 above.
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6/22/78, during a cooldown, the rate briefly exceeded the allowable je 100*F/ hour reaching 10*F/ minutes for about 3 minutes. In no hour period was
,.I the temperature reduced by more than 100*F.
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Repair Criteria j
1.
Explosive Expansion Process
.I a.
Update of testing and analyses to determine the effect of secondary side tube-to-tubesheet contaminants on the quality of the j,oint formed and on potential corrosion processes.
Testing to date indicates that tubesheet. corrosion serves to enhance the quality of the expanded tube-to-tubesheet joint. The tubesheets used in the qualification program were corroded to conditions simulating actual TMI-1 tubesheet material. In addition, a testing program is in progress using four tubesheet mockups corroded in varying degrees to provide a sensitivity
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study.
j The B&W Kinetic Expansion Report (Ref.17.4) discusses these programs in i
detail in socious 2. 2. 3, 2. 5.1.4, 2. 8.1, and 2. 8. 2.
b.
Update of activities to determine the residue left on the tube ID by the process, the usans of removing it, and/or the implications of not removing it (corrosion or other concerns).
Part of the qualification program involves a detailed evaluation of the
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materials present in the kinetic expansion process down to a level of one f
hundred part per million or the minimum detectable concentration.
All materials selected for use in the kinetic expansion process have been found i
chemically acceptable for use in the primary system. As part of. ou.r ef f ort to minimize the addition of contaminants, we have selected org~anic l
explosives to use during the process, and ordnance cord which contains the debris associated with transmitting the detonation energy to the kinetic j
expansion insert. Subsequent to kinetic expansion, pieces of polyethelene will be present, and some dark colored deposits on the tubes and tubesheet
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area. A detailed analysis of the deposits is documented as a part of the B&W Kinetic Expansion Report (Ref. 17.4, Section 2.7.1).
I Our approach to evaluating possible effects of this residue is two,f o ld'.
First, following iden'tification of composition of the debris we consulted bot's internal and external experts in the field of corrosion of reactor coolant system materials and attempted to identify by literature search and experience what, if any, deleterious effects we might expect from the residue which will remain in steam generator, albeit at very low levels. No adverse effects are anticipated.
Recognizing that this is not absolutely conclusive, we have implemented a long term corrosion test program using actual TMI-1 tube samples subj ect to the kinetic expansion process and the cleanup process planned for TMI-1.
These corrosion test samples have been placed in a chemistry environment comparable to that expected during the first cycle of operation. These samples are essentially lead samples which we anticipate will give us an I
indication if adverse effects can be expected to occur.
See Ref. 17.9, j
Chapter III.
t The OTSG tubes have been coated with Immunol prior to expansion to facilitate cleaning. Testing to determine the effectiveness of cleaning is
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discussed in Raf.17.4, Sections 1.4.3 and 2.7.2.
After expansion, felt plugs will be forced through each tube, and a water flush will complace the cleaning. Immunol and fait plugs have been evaluated for possible adverse chemical ef fects and found acceptable.
Any residual Immunoi left af ter j-cleanup will be removed by the normal purification system.
F c.
Expansion of cracked tubing to the tubesheet wall may widen or lengthen 1
existing cracks, forming a communication path between the primary coolant and the tubesheec metal. Updata testing and analyses to determine the implicati;ns of this, or summarize basis for concluding that there is no concern.
Work in this area has included evaluation of both the behavior of cracks undergoing expansion and the tubesheet behind an expanded crack.
I i-IF To date, one ' defect removed from the OTSGa has been expanded using 25 l
gr/ft and examined.
The defect was not visille to the naked eye on the CD of the tube prior to kinetic expansion whereas subsequent to kinetic expansion it had opened up approximately.03 r.ils and was visible under 6,3
(
power magnification.
There was no du: tile tearing as shown by metallography. It is our opinion that this defect was already 100% thru wall. (Ref. 17.4, Section 2.6.4).
Expansion and examination of several other cracks is planned.
In addition, several samples of sulfur-induced IGSCC grown in the laboratory have been expanded using 'DiI's procedures and examined, included were cracks less than 100% through wall. They also show no ductile tearing, but. a slight opening (fishmouthing) of the crack.
Since the repair process is qualifying 6" of the 17" or 22" long kinetic expansion which are free of significant defects, the presence of small. cracks below ECT detectability which have opened somewhat but not propagated circumferential1y or through wall should not affect the leak tightness or r
load carrying capability of the new joint.
L
- 2) There is a concern for boric acid attack af A508 steel if cracks do open as we currently anticipate. In reviewing previous testing we have uncovered i
sufficient information from B&W tube sheet crevice corrosion tests in the
'i late 1%0's and B&W model boiler tests for Florida Power Corporation in 1980 and 1981 to iodicate that we should not expect tube sheet degradation due to corrosion at operating temperatures and continuous exposure to borated t i water.
The testing that was done involved varied boric ceid concentrations and in the case of Florida Power Corporation testing approximately 6,500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> of operation. We have, therefore, concluded that the slight opening i
we may set in through wall cracks that exist in the tube above the bottom 6" of the kinetic expansion will not have any deleterious effects on the A508 steel tube sheet.
2.
Tube Plugging
[
What are the implications of continued cracking of plugged tubes (say due to L
sulfur on the crack tipe or thermal cycling).
If the tube completely parts, is the " free end" of concern due to vibration? Should tubes cracked outside of the I
tube-sheet region be completely removed or cut off below the lowest crack, and i
the free and " staked" to a support place?
l7 Tubes with cracks in areas of high cross flow will be stabilized. A summary of the criteria for using various kinds of plugs including those with
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17.9).
1.
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One action of the plan is to expand existing tubes into the tubesheet to i
provide a leaktight seal.
Do the qualifications for this expansion consider that the tubing is not new, e.g., trial expansion of sections of removed tube?
f_
It may contain incipient weaknesses not yet identified as defects. Expansion j. I might open potential defects.
Yes.
The tube material has been examined metallurgically and found to
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retain its normal characteristics in areas away from cracks.
(Ref. 17.4, l
sections 2.2.3 and 2.6.4).
In addition, evaluation of cracks which have
,.k been expanded shows no ductile tearing to enlarge the crack.
ECT h
calibrations indicated that cracks which will propagate under normal l
operating or iccident loads have already been identified.
Any incipient j
would propagate.
(Ref. 17.9, Chapter II.)
defects undergoing expansion are not expected to grow into the range that jl i
i 4.
Another corrective action is the plugging of certain tubes with a rolle'd
- n plug. Has consideration been given to the need to intentionally defect each
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plugged tube to preclude the existence of a pressure building up with the j
trapped volume inside plugged tubes?
4 The activity was considered but is not planned. All plugs, including the j
Westinghouse rolled plugs, are qualified for use with greater dif ferential j
pressures than those anticipated due to pressure buildup in a plugged tube.
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stabilising bars is addressed in the return to service SElt Chapter VII (Ref.
17.9).
. U 3.
One actica of the plan is to expand existing tubes into the tubesheet to provide a leaktight seal.
Do the qualifications for this expansion consider that the tubing is not new, e.g.,
trial expansion of sections of removed tube?
f, It may contain incipient weaknesses not yet identified as defects. Expansion 4
. I might open potential defects, f-Yes.
The tube material has been examined metallurgically and fcund to j
retain its normal characteristics in areas away from cracks.
(Ref. 17.4,
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sections 2.2.3 and 2.6.4).
In addition, evaluation of cracks which have been expanded shows no ductile tearing to enlarge the crack.
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calibrations indicated that cracks which will propagate under normal operating or accident loads have already been identified.
Any incipient defects undergoing expansion are not expected to grow into the range that would propagate.
(Ref. 17.9, Chapter II.)
4.
Another corrective action is the plugging of certain tubes with a rolled
- n plug. Has consideration been given to the need to intentionally defect each
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plugged tube to preclude the existence of a pressure building up with the trapped volume inside plugged tubes?
The activity was considered but is not planned. All plugs, including the Westinghouse rolled plugs, are qualified for use with greater dif ferential pressures than those anticipated due to pressure buildup in a plugged tube.
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