ML20126F002
| ML20126F002 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 12/15/1980 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Linder F DAIRYLAND POWER COOPERATIVE |
| References | |
| TASK-03-08.C, TASK-3-8.C, TASK-RR LSO5-80-058, LSO5-80-58, NUDOCS 8103050292 | |
| Download: ML20126F002 (5) | |
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Docket No. 50-409 LS05-00-058 JRoe TNosak RTEdesco Glainas ruW el d Mr. Frank Linder H
h General Manager Dairyland Power Cooperative NSIC 2615 East Avenue South TERA La Crosse, Wisconsin 54601 ACRS (16)
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Dear Mr. Linder:
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RE: SEP TOPIC III-8.C IRRADIATION DAMAGE, USE OF SENSITIZED STAINLESS STEEL AND FATIGUE RESISTANCE Enclosed is a copy of our evaluation of Systematic Evaluation Program Topic III-8.C Irradiation Damage, Use of Sensitized Stainless Steel and Fatigue Resistance This assessment comares your facility, as described in Docket No. 50-409, with the criteria currently used by the regulatory staff for licensing new facilities. Please inform us if your as-built facility differs froin the licensing basis assumed in our assessment within 60 days of receipt of this letter.
This evaluation will be a basic input to the integrated safety assessment for your f acility unless you identify changes needed to reflect the as-built conditions at your facility. This topic assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic is modified before the integrated assessment is cogleted.
4 Since rely, o\\
Dennis H. Crutchfield, Chief
.N eactors Branch #5
.v iconsing
Enclosure:
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December 15,?l980 4
pts Docket No. 50-409 LS05-80-058 Mr. Frank Linder General Manager Dairyland Power Cooperative 2515 East Avenge South La Crosse ~, Wisconsin 54601
Dear Mr. Linder:
RE:
SEP TOPIC III-8.C IRRADIATION DAMAGE, USE OF SENSITIZED STAINLESS STEEL AND FATIGUE RESISTANCE Enclosed is a cepy of our evaluation of Systematic Evaluation Program Topic.III-8.C Irradiation Damage, Use of Sensitized Stainless Steel and -
Fatigue Resistance This assessment co@ ares your facility, as described in Docket No. 50-409, with the criteria currently used by the regulatory staff. for licensing new facilities.
Please inform us if your as-built-facility differs from the licensing basis assumed in our assessment nithin 50 days of receipt of this letter.
This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your f acility. This topic assessment tray be revised in the future if your facility design is changed or if NRC criteria relating to this topic is modified before the integrated assessment is completed.
Sin rely,
.;y nnis M. Crutchfield, Ch' Operating Reactors Branch #5 Division of Licensing
Enclosure:
Cogleted SEP Topic III-8.C cc w/ enclosure:
See next page 9
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Dc wt, 15, 1980 7,r. Frank Linder-cc Fritz Schubert, Esquire Director, Standards and Criteria Division 4
Staff Attorney Dai ryland Power _ Cooperative Office of Radiation' Programs i
2615 East Avenue South (ANR-460)'
La Crosse, Wisconsin 54601 U. S. Environmental Protection.
' Agency O. S. Heistand, J r.
' Esquire Washington, D. C.
20460 M:rgan, Lewis & Bockius 1800 M Street, N. W.
Washington, D. C.
20036
- V. S. Environmental Protection Agency
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Mr. R. E. Shimshak Federal Activities Branch La Crosse Boiling Water Reactor Region V Office t
Dairyland Power Cooperative ATTN: EIS COORDINATOR P. 0. Box 135 230 South Dearborn Street L Genoa, Wisconsin 54632 Chicago, Illinois 60504 L
Coulee Region Energy Coalition Charles Bechhoefer, Esq.', Chair;.an ATIN: George R Nygaard Atomic Safety and Licensing 5:ard a
P. O. Box 1583 U. S. Nuclear Regulatory Coccission La Crosse, Wisconsin 54601 Washington, C. C.
20555 La Crosse Public Library Dr. George C. 'n:4rson 800 Main Street Department of Oceanography La Crosse, Wisconsin 54601 University of Washington Seattle, Washington 98195 U. S. Nuclear Regulatory Comnission Resident Inspectors Office Mr. Ralph S. Decker Rural Route #1, Box 225 Route 4, Box 1900 Genca, Wisconsin 54632 Cambridge, Maryland 21613 Town Chairran Dr. Lawrence R. Quarles Town of Genoa Kendal at Longwood, Apt. 51 Route 1 Kenneth Square, Pennsylvania 19348 Genoa, Wisconsin 54632 Thomas S. Moore Chairman, Public Service Cormission Atomic Safety ano Licensing Appeal Board of Wisc'onsin U. S. Nuclear Regulatory Commission Hill Farms State Office Building Washington, D. C.
20555 Madison, Wisconsin 53702 Ms. Anne K. Morse Alan S. Rosenthal,- Esq., Chairman Coulee Region Energy Coalition -
Atomic Safety and Licensing Appeal Board Post Office Box 1583' U. S. Nuclear Regulatory' Comnission -
Lacrosse, Wisconsin 54601 Washington, D. C.
20555 U. S. Nuclear Regulatory Commission Mr. Frederick Milton Olsen, III Resident Inspectors Office-609 North lith Street Rural Route!#1, Box 225 Lacrosse, Wisconsin Genoa, Wisconsin 54632 4
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SYSTEMATIC EVALUATION PROGRAM i
LACRUSSE BOILING WATER REACTOR i
i Topic III-8.C Irradiation Damage, Use of Sensitized Stainless Steel and Fatigue Resistance The safety, objective of.this review is to determine whether:the integrity of the interne.l. structures of operating reactors has been degraded through the use of-sensitized stainless steel.
The effect of neutron irradiation and f atigue resistance on materials of the internal structures was eliminated from the safety objective of Topic III-8.0.
in me randum to D. G. Eisenhut -from D. K. Davis and V. S. Noonan dated December 8, 1979. The memorandum concluded that operating. experience indicated that r.3 significant de2radation of the materials of the reactor internal structures hed occurred as a result of either irradiation or f atigue.
Furtner-more, the-Standard Review Plan (Section 4.5.2) does not address. neutron irra:1-ation nor: f atigue resistance of the materials of the reactor: internal structures.
As a result 'of incidents of intergranul.ar stress corrosion cracking:in pipir:
in the BWR system, special study groups were formed'by NRC'and industry to
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evaluate the cause, extent and~ safety implications of the use of ' sensitized stainless steel.in the nuclear steam supply systems.
The study groups identified the incidents with the recirculation system bypass ]ines, the core e
spray lines, and Che control rod drive return lines.
It was concluced that the problem was caused by a combination of high total stresses, sensitization of the.austenitic stainless steel in the heat affected zones of welds, and the relatively high oxygen content of the coolant.
J The NEC study group recommended an augmented inservice inspection program for stiiniess steel piping, more stringent monitoring of the leak detection system, modification of plant operating practice, and the use of. alternate materials immune to intergranular stress corrosion cracking.
The study group concluded in NUREG-0531, "Investige ion mid Evaluation of Stress-Corrosion Cracking in l
Piping of Light Water Reactor Plants," that intergranular stress corrosion cracking in piping would be detected prior to unstable crack growth because of the adequacy of the inservice inspettion program and the leak detection system.
Rea::cr operating experience has validated the leak-before-break concept of pipin; integrity, and, it was concluded, that through-wall cracks in the piping syste s would be detected before they presented a hazard to the health and safety of the public.
The regulatory position on the use of sensitized stainless steel in reactor internal materials is addressed in the Standard Reveiw Plan Section 4.5.2, "Rea:: r Intern: Materials." -The areas currently reviewed in the applican 's SAR are materials soecification and the controls imposed on the reactor coolant chemistry, fabrication practices.and examination and protection procedures.
The ma:erials specification should comply with Section III of >Se ASME Eoiler and Fressure Vessel Code and the components should satisfy the recommendaticns of Re;ulatory Guide 1.31, " Control of Ferrite Content in Stainless Steel Weld Petal" and Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless l
Steel".
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. The reactor vessel for the Lacrosse Boiling Water Reactor was designed, f abricated and tested in :cordance with the requirements of Section VIII of the ASME Boiler end Prissure Vessel Code,1962 Edition, and Nuclear Code Case 1270N, Stresses wers calculated by the methods described in the U. .
Cepartment of Co=e, :e Eu'letin PB-lS1987, " Tentative Structural Design Basis for Reactor Pressure Vessels and Associated Components". The primary stresses de not exceed the values :f Table UCS-23 and UHA-23 of Section VIII.
The reactor internal stru:ture is described in the Technical Section of the A::lication for Operating License, Ha:ards and Safety Analyses, and other 5:ecial Reports for tre Lacrosse Eoiling Water Reactor. The internal cer;cnents were designed :o provice support for the fuel and maintain the re:Jired ccnfiguration an: clearances during normal and accident conditions.
In accition, the internal ccm:onents provide passageways fcr the coolant to ccoi tpe fuel and means f;r a:ecuately separating :he steam from the coolan; water. The primary criteria for material selection for :he reactor internal ccmponents were the mscr,a-ical properties, the material stability and ccerosion resistance in the reactor environment.
The materials used for fabricatinc the reac::r internal comp:nen:s v.sre identified as Types 304 and~ 345 stainleis s:Sel, Inconel 600, and - ncr quantities of spe:ial ;urp:se alloys, such as Stellite.
Ex;erience has shown that n least three elements in ccmtination are necessary to cause cracking in sens'tized stainless steel c0nponen;s. ihese are r material susceptiLility, an oxygenated water environment, and a threshold total stress.
We assume #0r i.his evaluation that the La'rosse Boiling Water Reacter internal componen:s c:ntain sensiti:ed stainless steel in contact with an oxygen saturated coola-; wa:er environment.
However, the calculated stresses on the reactor components do not exceed the threshold stress values generally associated with inter;ran.lar stress corrosion cracking.
The threshold stress vaiues are near or grea:e-than the 0.2% off-set yield s:ress at temperature.
Further, in the react:r e vir:nment, stress reiaxation may occur due to irradiation and temperatu e effects.
The Monthly Operating Rep r:s, Licensee Event Reports and the BWR Ooeratina Ex:erience were reviewed #cr :he Lacrosse Boiling Water Feactor in craer to c:rreia:e reactor interna:.a:erial failure and the use of sensitized stainless i
steel in internal react:r com:onents. The events relate: to the failure of internal components are s.= arizec as follows:
Failure of fitti.gs in: : arts of the Core Spray Bun:le and Fuel Leakage Cetec:icn Systec was tra:ed to the use of cold worksd stainless steel fittings and par:s i thi system.
Tne mode of faihre was intergranular stress corrosion cra:k'n;.
The cold worked materia'.,cas replaced with fully annealed par s 1 : #ittings.
Similar events vere reported in Decemoer 1971 an: A:-i' '373.
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.c Roller liut Guide ' failure in the. Control Rot Drive Mechanism was reported in March 1969 and February 1976. Fracture of"the cast Stellite material cccurred.at.the end of the scram stroke, w'rere the higher mechanical stress is involved. Since scram capability is not impaired by this type f ailure, the failures did not contribute adversely to reactor safety.
Oegradation of a neutron source in November 1971 and several fuel rods in June 1973 and 1977, and May 1979 was attYibJted to intergran'Jlar stress
- ccrosion cracking.
Increased pressure in the neutron source from neutron raa:tions y<ith the beryllix and the fuel p21.let/ clad interaction in.the fuel rods resulted.in failure in the respe::ive events.
The defective ele ents were removed and replaced.
- uring remote inspection of the reacter intzrnals in June 1977, defects
,;ere observed in the peripheral shroud. Suisequent investication. showec
- hat the defects were caused by fatigue initiated in weld defects.
The shroud was redesicned and replaced with hearier gauge sheet material.
R ::Sclude from our review of the re;;r ed even:s tha; the integrity of the ria:::r internals for the laCr:sse 5 oiling Water Reactor was degraded throuch the c:e :f stainless steel which was subje:t to inte granular stress corrosien.'
The f ailures were detected by the inservice inspectior and testing procedures.
Tte f ailures did not adversely affect reactor safety, and were c:rrectec eitner by component remcsal and replacerren; or with aterial, immune :: inter
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grar.ular stress ctrrosion.
Tr.e inservice inspection program for the reactor internal cetonents is beinc c:.d.,:ted during the current interval to the recairements of $ection XI of ti,e A5"I Soiler and Pressure Vessel Code,1974 Edition, including Su rner 1975
%derta. The program is in Compliance eitn para;reph (g) vf Section 50.5sa c ~. CFR Part 50.
It will assure that the inteJrity of the components is rair.:ained during reactor operation.
Ke ::nclude from our review of the infor ation stbmitted by the licensee that t.e aterials in the reactor internal ccmpenents are ser.sitized and will be c:erated in an oxygen saturated water er.vironment. The incidents of. stress c:r-:sion crackinc are rare because the total st ess level is relatively low, r:: exceeding the 0.2% offset yield strenc n at gerating temperature.
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tre u likely event that intercranular st-ess corrosion cracking should occur, c-a:?s in the components will be detected by the inservice inspection and tisti.g procedures prior to component f a#hre. he conclude that the integrity
- ; e reactor internal components will ce assured by the inservice inspecti:n
- ;T conducted to the recuirerents Of Section XI of the ASME 5 oiler and F e25;re Vessel Code,1974 Edition, in:luding Sumer 1975 Addenda, in compliaice
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- aragraoh (c) of Section 50.55a cf 10 CFR Part 50. Further, we concludt
- a: intergranular stress ccrr:sion cra:<ir.g in the rea: tor internal componer:s i:
- a heard to the healtn and safety of the p;blic.
-