ML19247E433
| ML19247E433 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 04/10/1981 |
| From: | Basdekas D NRC |
| To: | Udall M HOUSE OF REP., INTERIOR & INSULAR AFFAIRS |
| Shared Package | |
| ML19247E428 | List: |
| References | |
| TASK-2.K.2.13, TASK-TM TAC-45198, TAC-45202, NUDOCS 8105110227 | |
| Download: ML19247E433 (2) | |
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C-r.itte: c i Inte? r and Insniar Affairs United 5tttes her:e C1 Representatives
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"n Yay 23,' ThS3 I hrote to you contarn'ng the safety implicat' ens of contr:1 s y: t a.: and.dyntsi: characteristics of nuclear p:ver plants. My c = a.nts then
-ere ir.' ended to dispute t.he official NRC position that " safe *y ryste=s will ciM; ate cent ci rys te::: failures at e.ny p:uer".
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in c.efer:nte 7 cf.f.htt letter. Such transien s can cause the reactor vessel to
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w i c:p:ble df catastr:phically fratturing a rt-actor vessal that has been exposed 4 n.u.r:n"finence ::rresponding : enly a few e il' P:ver Years Equivalent v
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- m:ven:er of such an a::ident, then,one should ask what are the chances 3
of it tak'n;.pic:e. Unfortunately, surh an t.ccident 's very likely and increa-in;1y sr. T.t is very likely because it =ay he caused by one or =re failures j
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mi he: aIreedy been reached. I believe that this level is pr:hshly n L.v
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>:-uer, that tran ient was nct :s severe as we ccn expe.:t en a reasenable
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beween the est'cetid vs.' the trasured values :f neutr:n fluence f:r t:e m ir.e Ytd ee rea::Fr vessel indicates a generic pr:ble that makes things
- cne. The results sf desize ry reasurements indicate the actut1 neutr:n i
fluence :: be scre7 3 times higher than that esticated in the t'.mine Ya.kee Fint) Iafety Analysds ?.ep:rt. 5bre:ver, as you ray recall, one :f the ceesures Ordered by HF. e.fter the TMI-2 a::fdent was to have all rea:t:r
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- n:t tur etff the ECCS :n=e it had been initiated. This eight be desirthic in ser:e cases of e::idents, but not necessarily in every case.
T r ver:: ling-trinsients, with:ut a large LOCA, the ::ntinued operation of EC:5 ::=peunds de a::ident by contributing to the :::1-dom process.
- and, st it;;r'2ndy, by repressuri:!ng the primary syste:.
He pressurihed 15N-.11 sh:.:k phenc ena have not been the subject :f e::pe-ricental w:rk by Se NRC nor the industry. lar nave the control syste=s.
an' their implicati6ns to safety been reviewed and analy:ed. Resa ::r.::ial sh:rt:::ings pose ~s5me cuestions en the effe:tiveness Of the regulat:ry pre:ess,' which y: Gay as easily es I ; nder, but the 1:r.ediate ::ncern is t: assure tha "'a y of operating plants. Freed with the realities that we are fa:d t:dayi and taking the a;proach tFat if we err, we sh:uld err in the dire:tien cit safety, it is e.p;trent t: r.e that' thos e Fi:7.'s with h';h ::pper ail y wesseis :r velds, thet have cperated for 4 FFYE cast be 7.hutd:w. until thincitter is res:1 red in the 'echni:z1 arena. It is ::n-s..ds,s a.i w..)
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to y:ur letter of re emberM,1950 en the safety implicatians of centr:1 systems in a fu rcnths, T.believe that this ratter is serious and pressing en: ugh, thet requi es e de;ision ncw. I believe that the C ::issi:n, with Congres-sienei assistance _and appreciation of the issues involved, will resp:nd
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If I can be of further assistance, please let =e kr.:v.
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Ce=etries L. Eastekas Retzt:r Safe:y Engineer W
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SUMMARY
OF THE INDEPENDENT CAPABILITY OF THE NRC STAFF AND ITS CONTRACTORS IN THE TECHNOLOGY RELATED TO PRESSURIZED THERMAL SHOCK The analysis of thermal shock to pressure vessels ~ requires the ability to identify a'nd analyze power plant system transients and then compute the pressure vessel transient temperatures, thermal and load stresses, and failure or fracture potential.
For the specification of the power plant system transients, the NRC steff has developed in-house capability to identify and analyze these transients for ranges of postulated plant conditions.
NRC contractors, including ENL, INEL, Sandia, ORNL and Los Alamos, have contributed to the development of rethods and codes for these analyses.
Experiments with model systems and components provide test data for code validation.
- For the analysis of the pressure vessel temperature transients and fracture potential, the NRC staff has developed computerized nethods independently from nore extensive and detailed methods being developed by CRNL under NRC contract. The NRC staff now also has the capability of running remotely the ORNL code on the ORNL computer with local print-out and plotting of the results.
Significant progress has been made at BNL in the measurement and calculational prediction of neutron irradiation of pressure vessels under NRC contract.
The methods are used to determine neutron irradiation for specific plants with various peripheral arrangements of fuel assemblies.
The HSST program at ORNL will include a new pressurized thermal shock test facility. This will provide a unique capability in testing pressure vessels to failure to determine safety nargins.
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7
, Enclosure 2 The NP.C staff includes personnel with wide experience and numerous oublications in the areas of systems analyses, materials applications, the effects of neutron irradiation on material properties, the effects of residual elements on weld properties and the application of fracture mechanics to-pressure vessel potential failure and fracture.
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