ML19260B535
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a G E N E R A L $ El.E CTRIC suctean powan SYSTEMS DIVISION GENERAL ELECTRIC OOMPANY,175 CURTNER AVE., SAN JOSE. CAUFORNLA 96125 HC 682, (408) 925-5722 p (c G nr ak<T gg g-3o { -19 2. - 3 I ~l U. 5. Nuclear Regulatory Commission (bO Division of Operating Reactors g.g g g Jgg Office of Nuclear Reactor Regulation CQdQ Washington, D.C.
20555 Attention:
Darrell G. Eisenhut, Acting Director Division of Operating Reactors 8 - % i-1 # 't. O U l 1
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Gentlemen:
SUBJECT:
ORNL CLADDING SWELLING AND RUPTURE DATA - 8WR EVALUATION
References:
1)
Meyer, R. O. and Powers D. A., Draf t Report,
" Cladding Swelling and Rupture Models for LOCA Analysis", 10/31/74 2)
" General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K", NED020566, January 1976 3)
Leonard, J.
E., et al, " Emergency Core Cooling Tests of an Internally Pressurized Zircaloy-Clad, 8x8 Simulated BWR Fuel Bundle", NEDE20231-PA, April 1976 4)
Levine, A. J., letter to C. F. Ross (NRC), "GE Loss of Coo 1&nt Accident Model Revisions - Core Heatup Code CHASTE 05", dated January 27, 1977 General Electric has conducted a review of the ORNL data ) 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 LOCA cladding swelling and rupture model ability to conservatively calculate peak cladding temperatures.
This letter is written to document our assessment of this situation as discussed with the NRC in conjunction with 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; 1519 048 7 91210 0 3 p 0
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GENER AL $ ELECTRIC U. S. Nuclear Regulatory Commission Page 2 Our review of the NRC supplied data leads us to the following obser-vations:
1.
The ORNL Multi-Rod Burst Test data on rod perforation and rod strain at perforation lie within the range of data used as a basis for References 2 and 3.
(Figures 1.B.2.4 of Reference 2 and F.2 of Reference 3.)
2.
The ORNL data does not inc'icate the presence of any trend which is inconsistent with the data discussed in Item 1 above.
3.
There are no new cladding strain perforation data (Figure 6 of Reference 1) in the range of BWR application (slow ramp data above 925*C).
In summary, the ORNL data does not affect the conclusions in References 2 and 3 over the BWR ranges of application.
Also, the approved GE LOCA cladding swelling and rupture model documented in References 2 and 4 is unaffected.
Furthermore, the sensitivity to changes in perforation criteria and strain at perforations for the GE design are small.
As a result of our review of this matter, GE has concluded the following:
1.
The models documented in References 2, 3 and 4 are in conformance with the requirements of Appendix K.
2.
All licensee technical specifications supplied by GE to ensure calculated peak cladding temperatures below 2200*F remain unchanged.
Clearly, this does not constitute a safety problem for the BWR.
We see no need to alter any BWR operating parameters or previously approved models.
I sincerely hope that this resolves any questions you may have regarding this matter as it pertains to the BWR.
Yours truly, M
R. H. Buchholz, Manager BWR Systems Licensing Safety and Licensing Operation RBH:mm/1432-1433 cc:
G. G. Sherwood R. Mattson, NRC 1519 049
"Babcodk &Wilcox
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P.O.Bos 1260, Lychberg,VA 2450s Tele #de (804)3St.5221-k
- Nowduber 2.1979 g
Mr. Barrell G. Eisenhut:
Deputy Director Division'of Operating Reactors Office of 1kiclear Reactor Regulation U.S. Nuclear Regulatory Ccaprission Washington D. C.
20555
Dear Mrf_,
Eisenhut:
Dn November 1,1979, the:MRC held a neeting with all of the hsel vendors and mangfof the operating plant utilities to discuss the imo' ct of a a
potentlaT revision of the cladding rupture edels used in LOCA avaluations.
BW pafticipated in that ceeting and responded fully to your inquiries.
This letter provides a written response to three questions raised by the NRC stilff i, the meeting.
T Questido 171 Do the fuel rupture and swelling inodels teet the' requirements of Appendix K of 10 CFR 507
.Respanse_
Thd BW evaluation model, as approved by the 3RC staff, $as been reviewd agiinst the criteria of Appendix K.
The BW evaluation model inchdes provisions which are conservatively imptemented relativeMo the potentici effects of clad swelling and rupture on a LOCA evaluation. The BW approach meets and exceeds the requirements of Appendix K,10 CRF 50.46.
Apper6 dix X in Section 1.B., " Swelling and Rupture of the Cladding and Fuel Rod Thermal Paramters," states:
- Each evaluationwdel shall include a provision for predicting cladding swelling and rupture from consideration of the axial temperature distribut%n of the chdding and from the dif fermr
~
iln pressure between the inside and outside of the cladding, poth as functions of tire. To be acceptable the welling and pture calculations shall be based on applicable data in such way that the degree of_ swellir,o and incidence of ruptu_re are t underestimated. The degree of swelling and rupture shah te ikken into account in calculations of gap conductance, cladding Aexidation;and embritticinent, and hydrogen vaneration.
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Sabcock&WWcox Mr. Darrell G. Eisenhut Movember 41979 The underitning has been added for emphasis. The B&W evaluatica inodel is based on appifcable data *and was approved by the tiRC in 1976. The model is docucented in BAW-10104A, "B&W's ECCS Evaluation Model" report.
The model works in the following way:
- 1. A stress versus rupture tegerature curve is mployed to predict the occurrence of rupture.
- 2. A ruptunrtemperature versus percent strain curve (or equivalent variable) is.used to predict local rod deforwation and appifed over a 3' segment for detensination of cladding tenerature.
3.
The local defomation determined from step 2 is treated as co-planar blockage across a fuel assewbly to establish the coolant
- channel flow area reduction.
4.
In conjunction with the reduced flow area, a one velocity head flow resistance drop is added to the flow channel resistance to predict post-rupture flow.
Adherence to Appendix K in the approved B&W evaluation rodel is provided as follows:
1.
B&W's existing correlations have 'been based on data which have been determined to be appitcable to the B&W design by engineering evaluation.
2.
The rupture criteria curve (Figum 1) over-predicts the incidence of rupture because it uses tesperatures for rupture that are 2ow relative to available data.
3.
The treatsent of the local percenc strain curve is an envelope over the applicable data set. The use of the maxiam strain our a full 3" setpent of the cladding over-predicts the cladding surface area for acta'l-water reaction and thus over-priedicts the reaction heating.
- 4. The assuntion of co-planar blockage is made in deteminino flow blockage up to an assembly blockage of 701.
If the co-planar assumption would create an assembly blockage greater than 70% the blockage is limited to E. These asstsaptions over-pmdtet the amount of flow channel blockage.
Applicable data refers to the availabie data set as of 1976 wnich was carefully evaluated for applicability to B&W plant LOCA ana?yses.
1519 051 g ; qg y o m M J LUAL
BabcockLWRcom Mr. Darnell G. Eisenhut
-3 Noveseer 2,1979 5.
It is expected that the one velocity head loss factor can, in addition, be shown to be conservative by a factor of approximately three on flow resistance.
6.
The gap conductance, after rupture, is dominated by radiant heat transfer and is, therefore, independent of the amount of strain at the ruptured location.
Camparison of BEM curves to the NRC curves is shown in Figures 1, 2, and 3.
ISEstated in the November:1; 1979, testing that w have not yet had the opportunity to assess the appifcability of the NRC data to our model ud assure the quality of the data. Ihwever, we believe that if the new data is assumed to be applicable to the ?&W evaluation model, it is the judgment of our technical experts that compliance with the requirements of Appendix K would be demonstrated.
For the fast ramp NRC curves. the applicable curves for large break LOCA, the B&W model over-predicts both the incidence of nJpture and the amount of flow chaseel blockage. The percent strain comparison shows that our approach predicts lower strain for local effects than the NRC envelope or upper beund curve. However, the NRC curve has been constructed to only account for th=
maximum strain at the point of rupture and is very conservative when applied over the entire rupture langth. The B&W strain curve, which is applieu over a three inch segment, results in swelling in agreement with the applicable test data and therefore does not under-predict the degree of swelling.
Question #2_
Do the plants meet the 2200'F criterion of 10 CFR 50.467
Response
As discussed above, the MW evaluation model meets the requirements of Appendix K.
The evaluation sedel has been employed for all BW operating plants and adherence to the criteria of 10 CFR 50.451s demonstrated Therefore, no BIM plant will exceed a peak cladding temperature of ZiWF.
nor the other criteria imposed by 10 CFR 50.46. However, the NRC presented curves _during the November 1st meetine which differed in saae aspects from those used by B&W. Without accepting the validity of the MRC curves, the remainder of this answer will establish the potential ispact of the NRC curves as if they were incorporated directly into the B&W evaluation podel..
The considerations of the curves igact is brokco into two parts:
(1) the impact on large break evaluations and (2) the impact on small break evaluations. We will first deal with large tireak LOCA's.
Lu 1519 052
Babcock &Wilcox Mr. Darrell G. Eisenhut Novecber 2,1979 For large breaks, the clad heating rates are predicted to lie between 0
15"C/s and 25 C/s at' the time of 7upture. The 25tc fast rmen curve would themfore be appmpriata for comparison to that utilfred in the M evaluation model. As shown in Figure 2, the BG curve lies below the NRC curve and would ~thus predict a higher incidence of clad rupture.
Excepting the peak in the NRC curve at 1.6 KPSI, the PRC blockage curve is either in agreement with the B&W curve, at high stress, or below the ESW curve at moderate or low stmss, Figure 3.
Thus, the PRC curve would predict less reduction in flew and greater cool-ability after rupture.
This statseent applies during blowdown. During reflooding, the FLEOf7 data has been judged to predict heat transfer conservatively for blocked or untiocked corws so long as flooding rates are steve 1 1p/sec. Statements to this effect am contained in Ardendix K of 10 CFR 50.46. BWs fToedig cates
. are above 1 in/sec for all evaluations, well past the time of' peat ddffag temperatures. For some evaluations where the peak occurs higli in the core, a lower than one in/see flooding rate can occur for a short time after the time of peak cladding temperature and before con quench. It&W u;es ( conservative steam cooling model, including blockage effects at this tinei As sentioned,
.the degree of blockage is over-prettfeted because the blockage occurs prfor to peak c adding temperature and, tinrefore, the present siedel R otewevathe.
The local burst strain curve of the new MRC data is above the curce used by BW fn its evaTuation sedel (Figure 2).
Tnese curves are derived (ttna measurenents of maximum local strain at the point of rupture, b M's model, the strain from this curve h applied uniformly for a full three inches along the fuel pin length. Realistically, this raaximun strain probably exists over no core than about one inch axfatty, with reduced strafn eftewhett.
It
'is out $u6gment that the higher st. rain values of the PfRC curve, if used with i more realistic axial model of. the rupture effect, would predict cladding swell and exposed surface area not significantly different frca the existing 8&W evaluation model.
Should the higher burst strain curve of Figure 2 actually result in greater total clad dislocation, it is not likely to increase peak clad tescerature significantly. Excess strain would manifest itself mainly in higher heating rates due to metal-water reaction, as'~ discussed below:
1.
The gap wouldTincmase. Because.gep heat transfer is dominated by radiant effetti,little change in beat transfer would occur.
2.'
More surface area would be exposed for metal-water reaction, thss slightly higher heat rites and more oxidatfon'weuld be caicalitted.
The iveact of_the higher heating rates is inconsequential becape:
a.
BW predicts 'upture to occur durhg blowdown when coolant is available to remove the Imat.
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Rmhew&AWiiCOX fir. Darrell G. Eisenhut
-5 November 2,1979 b.
Ithen rupture occurs the cladding in the vicinity of the rupture cools due to the reduction in gap heat transfer.
The peat cladding tecperature generally does not occur at a c.
ruptured location on the fuel.
d.
for those few times when the ruptured location sets the peak cladding te@erature, the driver for cladding tenterature is mainly fuel pellet energy as opposed to metal-water reaction, and, thus, little impact would result. In these calculations, which fom the basis for LOCA limits at the 2-and 4-foot core heights, some minor impact on peak clad temperature might occur.
Our judgment is that pis will not cause an increase in peak tempt rature past 2200 F.
The rupture strain nrve used in the B&W evaluation model is concluded to be approximately equivalent in application to the NRC peak strain curve were the NRC curve to be applied realistically with strain determired as a function of axial distance from the rupture.
It is further concluded, however, that application of the higher NRC peak strain curve over the axial iength assumed in the B&W evaluation model would not result in peak cladding tempera-0 tures in excess of 2200 F.
For small breaks, the NRC slow ramp data would be appropriate. B&W small break analysis does not generally predict any claading heatup for small breaks because no core uncovery is predicted. For a very specific set of break sizes and under full Appendix K assumptions, a small amount of core uncovery and resultant clad temperature excursion has been predicted.
For the worst case, the peak cladding temperature is 590 C and the stress fs below 8.4 KPSL 0
Tgis temperature and stress corrbination is approximately 150 C below the O C/s rupture stress NRC curve. Therefore, no rupture or swelling will occur and the use of the NRC curves would have no impact on our small break predictions.
The above coaments apply to all operating BAW plants regcrdless of design because the implicatfcas drawn ore taken from the evaluation model irtifch is applied uniformly regardless of design.
Question #3 If the response to Question #2 is "no", Afine proposed ir.terim actions the affected plants might take.
- Besppple, In view of B&W's response to Question #2, this question is not applicable to plants utilizing the BAW NS55.
1519 054 p-g egoy7 z um
Babcock &Wilcox Mr. Barrell G. Eisenhut Novenber 2,1979 In conclusion. B&W believes 1.
that its present codel is in compliance with 10 CFR 50.46 and Appendix K, 2.
that a unifonn inclusion of the NRC proposed rupture and strain curves would have no imact on our evaluation of LOCA, and 3.
that even the most conservative inclusion of the NRC curves would have little igatt and not result in predictions of cladoing teraperaturns in excess of 2200 F.
In view of the above, we believe that the new data does not change the contiusions of the previously completed safety analyses for the operating plants utilizing the BFR NSSS.
Ve ly yours, yf f;fN/
f James H. Taylcr
/
Manager Licensing Nuclear Power Generation Division JHT:nw Attachments cc: Mr. B. 6. Borsum 1519 055
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Tek r 9 9297 e
1000 ProstwJ Mdi HOSO Wandtor, CrJnncCllC41106095 POWER SYSTEMS Novmber 2,1979 LD-79-064 Mr. Darrell G. Eisenhut Assistant Director for Systes and Projects Division of Operating Reactors U. S. Nuclear Regulatory Comission Washinsi.on. D. C.
20555
Subject:
Fuel Cladding Swelling and Rupture Models
Reference:
Letter, LD-78-069, A. E. Scherer to D. F. Ross, dated September IB.1978.
Dear Mr. Eisenhut:
At a November 1,1979 meeting your staff requested that, to support our operating plant customers, Combustion Engineering (C-E) address several NRC concerns on rupture straia and flow blockage. To obtain some perspective on this problem, however, it should be noted that licensing evaluation models - per Appendix K Criteria - are inherently conservative.
Results using more realistic ECCS Evaluation Models indicate that the cladding temperatures are 50 low that no flow blockage will occur even during a postulated large break LOCA.
W1th this perspective, we believe the evaluation of the impact of 3 0ur-strain and flow blockage concerns is an ad:ninistrative rather than a safety concern.
It should be noted that the following evaluation applies to C-E designed tiSSS's utilizing C-E fuel and ECCS model.
The new NRC curves of rupture strafn and flow blockage have been compared to our revised flow blockage andel submitted vi3 the referenced letter to address the fellowing two questions:
QJESTION #1 Does C-E's Flow Blockage Model Heet Appendix K?
ANSWER fl C-E believes that its Flow Blockage Model does meet Appendix K.
The basis for this reply is presented below.
One elment of the C-E ECC5 evaluation model is the elment which simulates the consequences of fuel rod defomation. This element includes calculation of a number of submodels. Specifically, geometry of an individual fuel rod, geometry of coolant flow channels, flow diversion and recovery caused by fuel rod defom-ation, and changes in fuel rod heat transfer characteristics caused by fuel rod defomation. The combination of the component models which descrit>e these in-dividual phenomena is referred to at the flow blockage / heat transfer element.
D**]O *D
~ 'W }f 3 c o n d R a n-1519 059
' Mr. Darrell G. Eisenhut '
In response to discussions with the NRC Staff in 1978, a revised rupture strain model was developed by C-E.
In addition, improved models for flow blockage and steam cooling were developed. These alternate models have been sutnitted to the NRC Staff for approval. The C-E alternate rupture strain model has been developed based on recent out-of-pile and in-pile data.
This alternate rupture strain rodel is a data upper bound and, when depicted in a rupture strain vs. rupture tcepera-ture graph exhibits a rupture strain " hump" for rupture temperatures below about 0
900 C and is similar to the latest NRC rupture data presentation.
In particular, for the heating rate range (about 2 to 100C/sec) and rupture temperature range (about 8500 to 9500) typical of C-E NSSSs, the C-( revised strain / rupture curve natches very closely the most recent data envelope presented by the NRC.
QUESTION #2 Do C-E Plants Meet the 22000F Peek Cladding Temperature (PCT) Criterion?
ANSWER #2 Including the data presented at the p.eeting of November I,1979, C-E concludes that all plants operating with C-E fuel continue to meet the 22000F PCT Criterion.
The clad rupture conditions for current C-E operating plants are grouped in the range of 850 - 9400C which fall near the valley of the strain vs. rupture tertp-erature curve of Figure 1.
Those C-E plants which are calculated to experience fuel rod rupture during the reflood period of a LOCA, exhibit the greatest sensitivity to changes in rupture strain and/or flow blockage due to the Apperdix X requirements for steam cooling and blockage for the less than 1 in/sec reflood period. The Systen 80 sensitivity study, previously documented, is characteristic of these limiting analysis conditions.
The results of the System 80 sensitivity studies show that PCT calculated with the revised flow blockage / heat transfer model results in slightly lower FCT tien the present flow blockage / heat transfer modeL In addition, the results of a study show that increasing the degree of flow blockage from 60I to 80% only increases the PCT by 400F.
We believe that the results of these studies are iridicative of the results from C-E NSSS's cxhibiting reflood rupture. Those C-E operating plants, which have been calculated to rupture during the blowdown or refill periods using the approved ECCS evaluation nodel, will not be adversely affected by the changes in strain or blockage.
These cases are all PCT limited during the reflood period.
As a result of the above assessment, we believe our Evaluation Model with the revised flow blockage / heat transfer model meets Appendix K requirements. Further, we conclude that the 220rPF peak cladding temperature criterion is met.
If I can be of any further assistance on this matter, please contact me or Ms. J. M.
Cicerchia of my staff at (203)683-1911, Extension 2595.
Very truly yours.
COMSU5 TION EPr3 NE NG, INC.
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