ML19260B583
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E@ NUCLEAR COMPANY,Inc.
2101 Horn Rapids Road P. O. Box 130, R:chland, Washington 99352 Phone: (509) 943-8100 Telex: 32 6353 November 2, 1979 Mr. Darrell G. Eisenhut, Acting Director Division of Operating Reactors Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Dear Mr. Eisenhut:
Exxon Nuclear Company (ENC) has reviewed the information presented by the NRC staff at the November 1, 1979 meeting regarding recent rod burst test data. We currently conclude that the data presented are properly represented by the fuel models used in the approved ENC Evaluation Models. Thus, the analyses performed by ENC for both PWRs and BWRs, all of which demonstrate that the criteria specified in 10 CFR 50.46 are satisfied, continue to be applicable. An expla-nation of the basis for this conclusion is given in the Attachment to this letter.
Sincerely,
' G. F. Owsley, Manager Reload Fuel Licensing Exxon Nuclear Company 1519 062 AN AFFILIATE OF EXXON OORPoRATION 7912100407
ATTACHMENT e
EFFECT OF NEW R0D BURST DATA ON ENC r'UEL MODELS PWR Blockage The new NRC data suggests that flow blockage may be ramp rate sensitive.
The ENC flow blockage model, while not ramp rate dependent, does bound both fast and slow ramp rate data, Figure i shows the comparison of the ENC flow blockage model with the new NRC correlation for high ramp rates.
In the rupture stress range of 3 to 6 kpsi applicable to ENC PWR reload fuel, the ENC flow blockage model is very clearly conservative and hence is not at issue.
Figure 2 shows the ENC flow blockage model versus the NRC correlation for slow ramp rates. Also shown is the new multirod blockage data for slow ramp rates.
It is apparent that in the rupture stress range of interest the ENC model bounds the data and hence fully satisfies Appendix V criteria. The new NRC correlation is overly con-servative with respect to the data.
In part, this is a result of convo-luting the blockage versus stress from the stress / rupture temperature curve as opposed to directly correlating blockage versus rupture stress.
PWR Ruoture Strain Figure 3 shows a comparison of the new strain data to ENC's strain model.
In general, the ENC model bounds the new data for both slow and fast ramps in the rupture temperature range of interest, 880 C to 950 C.
PWR Rupture Temperature The rupture temperature data presented by the NRC suggests a slight tempera-ture ramp rate effect in the range of interest. The ENC rupture temperature model is bracketed by the new data. The small differences over the range of applicable ramp rates are not significant.
BWR BWR conditions during rupture typically have slow temperature ramp rates with rupture stress in the zero to 2 kpsi stress range.
The new slow ramp data indicates rupture strain in the 20 to 30 percent range which is consistent with the ENC BWR rupture strain and blockage model.
Furthermore, the new rupture temperature data is also consistent with the ENC BWR rupture tempera-ture model.
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Todophone 60 306-991 sus ee-eee-eeee YMEEEATQWN ELECMR 4MEPMY m.,,s...o.,
. - mm November 2,1979 Galted States Wocieer Regelstery Casseisaies w::'n=g=, 9.C. 30SSS Attenciast Office of Basleer Reactor sagstaties Mr. Darrell Eisenhet acting Sizester Rafarences (1)
License No. BrE-3 (Doclast Es. 90-19).
(2) 41addiag Suelling and Rapture 18edals," draft dated 16/31/79 by D. A. Fausrs and 3. 9. Meyer.
(3)
TABC letter to USanc, Re "%seswesed Core XIV Zsicial fuel pia Fill Pressure," dated Angsat 21, 1979.
Dear sirs Sebjests Evaluation of Cladding, Suelling and Empture Models this letter is la respomac to concerns raised by yese etsif daark,g a meeting on sevesbar 1,1979 la tukick a draft report was presested ob' A cameniand correlatioon reistias to claddias behavior==ame LOCA cauda.tions, Referemos (2).
At that tias, soggestians were made by the MC staff stat suisting carrelatiees used la I4Ca analyssa asy not enveleye ser data seleting to elad defermetion and rupters. These correistions inslades (1) cladding burst temperatux*/ stress, (2) claddias burst temperstare/strais, and (3) redacties in flau ares / screes.
In this letter, theos concerns are disemesed with referessos to the 14CA calostations Yaac has p L-cr' to =,_G eperatiam of Yankas Eaws dering the current cycle.
YAsc at preseet perfesas a borney eensitivity study for LOCA calm =*4:z for the Yankee Rome mactor to defies a 11asar best generation rate limit as a functiam of core hermer. Its method employed and thm pois.:s analysed for the present eyeratius cycle (Cycle 14) are decessanted la Esference 3.
(1) claddi== a=est Temeeracare ver=== Easiaseries sees ser we The LOCA annipis for Yankee Base utilisee the IREM burst temperature curve to pundict incidenee of rapture.
In all saalyses performod, no rupture is calculated to occer. The specific analyses as described in Referomoe 3 for the curress feel in the Tankee Rowe reactor have been saamined to deterzien the estus, calculated cimeding stress-temperetare points. Of sixteen cases sundmad, only one case lies above the O'C/ses serve of Referenso (2) and all cases are in the reage of hoop stress below 4 kysi.
11ae single case above the OaC/sec cos9ee lies withis tbs 14egf.,e,,,,,,
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Office of Musleer Reactse Regulassen Boeember 2, M Paga f This examinatalam demametrates ehet applicacias of ramp rate dependent curves prssented is tim draft report, Referesca (2), will met affect the results of our calostatiens far Yankaa Ross with respect to burst temperature /psweemre.
(2) C1 "s marst serais v.m 7 - : r e
To evaluate the implicatica of ekie correlation la the dcast espect, Seferemos (2), the burst straias surve for slos ramp rates (that thish is applierbie to Yankee Rams) une subecitated for the Y4BC correlaties.
Calculations were redema far the fresh Seal assembly case at abanc 6800 Md/Te termuy to zapresent our surrent operating condities.
The remedts abow that peak claddlag ature does mec chases significantly
(<3D*F) and resales well holes (3), Local _ Flow Blar+see woraus Remianeri== Boop str===
To ovalente the implication of this correlation in the draft report, Refonneos (1), the Cangesite Ammeubly Flow Shkage curve see sabetituted far the D SC ascreLetica. CeJ.oulations mere redens for the freak feel ensembly case et chest 6000 JEld/Te hermey to represent the meet limiting beadle is our carrent operating eendition.
No raptars was predicsed to occur in eitkar the wriglasl coloulation er la this revised smalysis.
It was establiebed that <his parameter had oc infinance on the peak clad temperators la the. ass analysed.
In the time period availabic TABC hau reviewed its 123 emelysis and resulta for the Yankee Rene plant in ilight of the information presented im Reference (2).
Specific celos1stiano hawe been pr"ormed at a core buswup coesidered to be representative of the surrest orperstlag condition of the The rseutta empport contimmed operation and ensure safety and health reactor.
of the pad >11c.
De will coattimes to moviser the data, correlations, med analysis in the draft report and their implicatione with respect to the Ypakne Ross reacter.
Teaken is comeermed at the natuce of the pness involved in elevating this Leeue to the starm of amadated insraec coepoese eeder clas implication of immediate plast staatdoom.
TMSD was prevised a copy of the draft report, Reference (2), so Thursday, November 1 during the meeting in Bethesda.
(Iacidaatly, invitation to this was extended informally via telephone late in the afternoon on ilmanasday, October 31.)
The draft tapert ses admitted, by the staff, to lack any peer causeast, but was presented M the basis of imanadiate coaccra. Altbewab evaluatian e' the report was la progrees, TE5D was met advised ess;ii
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Offles of Ntalear Rameter Regelation Newmmber 2, 1779 Page 3 4,;n-Mely 11 00 a.m. en Frie
,v, November 2 that a response in writing was espected by this afternoon, Develegament of a written reopease in this time fresne pieced a severe harden on the orgamination and presluted inweatisatics of several relevant isenes including assessmeet of the appropriateness of the emperim*=1 data base and its application.
The feet in there are areas of phamanasological ameertalaty in accident aselyses.
This is,a
.;dly tery conservative anstyees are Wred.
As eer understanding of such phoemamas imeresses, we obs11 cometantly be faced with avslanting existing appresehes sad ssedifying thesq quite possibly to sumove umwarrested ceeservatisme. In any case, it is Tankee Atomie's peettion that this preesse of eyelastics sad authed insprovement most proceed carefully and preeently. This should not be a preessa of immediata implementation of the MRC's leemst viewpoint witbewt reasemed testesiemi review.
If as accept such a preessa, it is Taskee's eyimien that continese safe operaties of unselaar paper pisata will be jeopardised If you have any questicas ragsrding the enehmical costant of this letter, please feel iese to easteet Dr. Aasssf Bosaim or Dr. Stephen Sebalta of om?
Bastaar Engineering and Develegament Department.
Very truly years.
TAMME ai1EGC EutOfB1C 00MPA1rf YNP,1,^
- s. a. emeneebersh samloe Vice Presideae Bgr/mma non oN RX
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1519 069
GEN ER AL @ ELECTRIC nucteaa eowna SYSTEMS DIVISION GENERAL ELECTRIC COMPANY,175 CURTNER AVE.. SAN JOSE, CALIFORNIA 95125' MC 682, (408) 925-5722 i
U. S. Nuclear Regulatory Commission Division of Operating Reactors Office of Nuclear Reactor Regulation Washington, D.C.
20555 Attention:
Darrell G. Eisenhut, Acting Director Division of Operating Reactors Gentlemen:
SUBJECT:
ORNL CLADDING SWELL AND RUPTURE DATA - BWR EVALUATION
References:
1)
Heyer, R. O. and Powers, D.
A., Oraft Report,
" Cladding Swelling and Rupture Models for LOCA Analysis", October 31, 1979 2)
" General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K", NEDO 20566, January 1976 3)
Lecnard, J. E., et al, " Emergency Core Cooling Tests of an Internally Pressurized Zircaloy-Clad, 8x8 Simulated BWR Fuel Bundle", NEDE 20231-PA, April 1976 4)
Latter, Levine, A.
J., to D. F. Ross (NRC), "GE Loss of Coolant Accident Model Revisions - Core Heatup Code CHASTE 05", dated January 27, 1977 General Electric has conducted a review of the ORNL data (Reference 1) presented by the NRC staff in the November 1, 1979 meeting.
We understand that in the NRC's view the data raises questions on vendor's ability to conservatively calculate peak cladding temperatures.
This letter is written to document our assessment of this situation as discussed with the NRC during the November 1,1979 meeting.
Based on the information presented at the meeting, we understand the staff's concerns lie in the areas of:
- 1) fuel rod perforation criteria;
- 2) fuel rod strain at perforation; and 3) flow area reduction.
Of these concerns, only fuel rod strain at perforation relate to the BWR.
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