NRC Generic Letter 1979-66

NRC Generic Letter 1979-066: Information Regarding New Fuel Cladding Strain and Fuel Assembly Blockage Models and Compliance with 10 CFR 50.46

ML031320434
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Nine Mile Point, Palisades, Indian Point, Kewaunee, Saint Lucie, Point Beach, Oyster Creek, Cooper, Pilgrim, Arkansas Nuclear, Three Mile Island, Prairie Island, Brunswick, Surry, North Anna, Turkey Point, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Duane Arnold, Farley, Robinson, San Onofre, Cook, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Fort Calhoun, FitzPatrick, Trojan
Issue date:	11/27/1979
From:	Eisenhut D G Office of Nuclear Reactor Regulation
To:
References
NUREG-0630 GL-79-066, NUDOCS 8001070276
Download: ML031320434 (32)

.V _. I- ---L-?UNITED STATES NUCLEAR REGULATORY

COMMISSION

WASHINGTON, D. C. 20555 November 27, 1979 GI-- (LETTER TO ALL OPERATING

LIGHT WATER REACTORS)Gentlemen:

Letters, dated November 9, 1979, were sent to all the licensees requesting information concerning the new fuel cladding strain and fuel assembly blockage models and compliance with 10 CFR 50.46. Subsequently further information was provided by the NSSS vendors and fuel suppliers regarding the impact of cladding heating-rate dependent burst temperature effects on rupture time and rupture strain/blockage and consequently on calculated peak cladding temperatures.

The vendors and fuel suppliers supplied additional information by letters to the staff.Copies of these additional letters are enclosed.

This new information should be used in preparing your response to the November 9, 1979 request.incerely, Darrell G. senhu ting Director Division o perating Reactors Enclosures:

1. Letter from 2. Letter from 3. Letter from 4. Letter from 5. Letter from 6. Letter from Babcock & Wilcox, dated November 20, 1979 Combustion Engineering, dated November 16, 1979 Exxon Nuclear Co., dated November 16, 1979 General Electric Co., dated November 16, 1979 Westinghouse Electric Co., dated November 16, 1979 Yankee Atomic Electric Co., dated November 20, 1979 C?)e) -6-1-7 Baboock& icox v -24 PO. Rox3260. 4=?:*s, Va 24505 Terephioxe:

MU 384-51 13 November 20, 1979 Hr. Darrell G. Eisenbut Acting Director Division of Operating Reactors Office of Nuclear Reacter Regulati" U.S. NucleUr Regulatory Commission Vashington, D.C. 20555 SubJect: ClAdding Svefling and Rupture Models fcr LOCA Analysis Dear Mr. Eisenbutt On Wovebear 14, 1979, Kr. R F. Denise of the Division of Systeas Safety contact I&W with regard to the Burst Temperature Curve approved for nze by BE&W in LOCL analyses.

li. Denitse re-quested BW to considcr the effect of ruloying the Staff's ramp rate correzltion.

a contained in Draft KMMO 0630 to deteridne the Burst Temperature Curve for use in WC& analyses.B&' has exained the ramp heat up rates calculated prior to rupture for B&W WSS syste= which have either OLs or CFs granted under 10 CW 50.46. (Documented in EA1-10102, Rev. 2, EW-10103AI, Rev. 3, ind EA-1010S, Rev. 1.) Ilterpolating from the Staff's r heat up rate versus imp stress and failre temperature referenced above, BSW has found that the Staffis correlation predicts the fuel cladding to rupture at the same or higher tamperatures for al1. cases, except the 4-foot coe elevation for the £77-Fuel AssSemby raised-loop pnt (SAW-10105, Rev. 1). The razp rate prior to rupture fI this case is approxinately

12 C/s, whle the ertrapolation of the NkEC curves to the EW Burst Temperature Curve at tbat &ae stress Indicates

& 22 Ca hte t up rate. SW has estimted tbat the effect would be an earlier rupture, and, thetere, additional oxidation due to metal-water reaction, rulting in an increase of appr"imtel

6O1 In the peak cladding tupeerature (PCT).The original analysis showed a peak cladding temwrature of 2073F. The addition of 80? wod result iu w peak of 2153'F and not violate tie requirements of 10 CFR 530.4. Since the issuance of -1l0105, Rev. 1, B his idntified further con-servatisms which wmount to a reduction in peak cladding tempera-ture of approximately

3W .Therefore, if the evaluation vere The 9Obcok&VWikt Coz-V8V I W4tstt~f 1867 t Rabcock & Wfl=o Mr. Darrell Eisenhut November 20, l979 to Include these further conzervatims, and tie MC ramp rate correla-tions employed, we wioud expect a peak cladbiuS temperature intreate of about 50)F (2123V7 peak) with no difficulty in demonstrating om-pliance to 10 CFR 50.46.In suzaary, BW bAs examined the effect of the use of the Staff's ramp rate correlation as requested and found the calculated PCT to be either unebanged at lowered as a result except for the one case noted above. If there are any questions concerning this response, please call me or Resry Bailey (Ext. 2678) of my staff.Very truly your,.3anes R. Tylor Manager, Lkensing JT/lc zz t Nd Oa 'AQN 6 ,wv e'.A~oV I ..

ComhLEDfl nrinrr;sn3.

o TI.cl~e 9 229 1W0 PIO5PeCl Hiii RoOi Windsm, connemicul:

W095 POWER SYSTEMS Roveber I6. 1979 LD-79-067 Mr. Darrell C. Elsenhut -Assistant Director for Systvs and Projects Division of Operating Reactors U- S. luclear RegulatorY

cu'issidO Wshingtnn, D.. L 20555 Subject: Fuel Cladding Shelling and Rupture Nodels Reference:

Letter LD-79-064, A. E. Scherer to 0. G. Eisenhut.ste November 2. 1979 Dear Kr. Eisenhut: The rferenced letter prmi4ded Conbustion Enginfeerig's (C-E) response to seyeral NRC concerns regatrdino rupture strain ard flow boage.. Su-sequent to receipt of the letter, additional questilons armse concerning the impact of heating rate dependent burst tonperatre effects on rupture time and rupture strainfblotkage arnd ultimately an peak cladding tem-per-ature (PCiT) in the analysis of a large break LO£A. The following evalua-tion of the potential impact of heating rate dependent burst taterature on C-E'S licensing calculations is provided in support of our oparating plant custineri.

The C-E rupture temperature model does not have heating rate dependence.

For the heating rate range of C-E operating plants, 2-10yCfter znd using the ORL mrodel reccmended by the Staff. heating rate effects would lower predicted rupture temperatures by 25-750C. The resulting lower rupture tMrperatures due to low heat rate effects produce earlier rupture times, (2-20 se-wnds earlier).If rupture occurs after the tine of <1 Wnsec ref loIW. degraded heat tran&-fe on the rupture Woe ard above (as required by Appendid K) is invoked at the time of rupture. Earlier rupture times could lead to higher re-flood PC in this case because of the earlier i silipirentatAor.

of degraded heat transfer.

However, all C-E operating plants experInze c3ad rupture prior to the time of el in/sWm reflWd and therefore the initiation of degraded heat transfer would not be affected by lower rupture topiatures.

Earlier rupture times during the blawdown or refill periods mnay alter local beat transfer *ncmentarily, thrTouh wp conductance or radiaticn enclosure effects. Bowever, if the PCT occurs during late reflood its imPact on PIT woUld not be significant.

r Dc Orrell G. Ekernhut-Z

Lower rupture temperatures due t low I e3t rate effects may produce higher rvpture strains and blockages.

The effect of increased rupture strain and blockage was addressed in the referenced letter tg the Steff.The results of the previously discussed System 80 sensitivity studies show that PCT calculated witb the revised flow blockage/heat transfer model is slightly lower than PT calculated with the present flowblockaagef heat tran5fer model. In addition,, the results of a study show that in-creasing the degree of flow blockage from 6DO to 8DX only Increases the PCT by 40 0 F. &ised on these results, we conqlude that all C-E operating plants continue to coinply with the ZZOCOF peak cladding teperature criterion, Including the effects of increased rupture strain/blo-kage.

The above discussions indicate that the reported PCT for all C4 opeating plants would not be siganficantly affected by a haating- rate dependent rupture temperature model. The magnitude of the effect on PC would be no greater than effects observed in the System 80 sensitivity studies using the C-E alternate rodels. In facts it is expected that uting- rL-vised flow blockagefheat transfer models with or without a heating -ate dependent burst tm aperature model for the analysis of C-E operating plants h-Duld produce lower PCT than presently reported values. C-E therefore believes that our Evaluation

%Idel analysis with thf revised flow blockage/heat transfer miodl meets Appendix K requirements and the ZZaODF peak cladding tepperature criterion.

If I can be of any further assistance on this rmitter, pfiase contact uze or Mls. 3. H. Cicerchia of uy staff at (20)3S1-1911.

ExtenSon slo 5.Very truly )ours, CtiBWST1It BEIE~ThERINr INC.Licensing tBanager AES:d-ag lk*CtJ UCLt&Ahb :IYIne:.1m tC),ll L0 lute flt Horn hopids AD.,d 0. V>. &&x fjO, P~h0"ffJ~trl.

li'st~t~fo g3t72 PAiinr f5019 94.Y R1.0 rale 32ti3 Novotnbpe

1t, 1q79 Mr. barrell G. tisenhut.

Acting tirector bivision of Operating Peaictors Office of Nuclear Rpart.ror Regulation U. S. Nuclear Regulatory Conmnmission Washington, D. C; 2O!55 keferenrce ti: MNt letter frnmi G. r. Owtlpy to h. A.. tist.albut dated Nnvft*p&tt

4, 19I/9.betit Har. Eisenhut: A!s iaquested by your staff on November 13, 1979, tMt hp! cnmpleted Ah additional review of the licensing imapct of thr: rsriujed NRC rupture/blockage model with pattLuler emphasis on the impact of the NRC tempera-ture-ramp-dependent rupture temperature curves. Tthi .Peview SupportS the conclusion of ENC's earlier anal yte (Aeferehrt!

1) that there is ho adverse impact on licensing lint ts for plants AnhAled by ENt modelt ttfo ute of the NRC rupture/blockage model.the DC Cook analyses reported in tPftPvPnhr.

(1) UP'ed the to:mrole.tr:

Lt'ruLso'stl NRC rupture/blorkage model including the temperature-tImp-dependent tuptuiC temperaturv, rupture strain And flow blockage cbrrelatinF..

Thin., the temperature ramp rate effects had been included.

The teqserature rtip rate dependence in the NRC model is such that the difference between the predicted rupture temperature of the HRC and MNC tmdels Is gredlest for the 6lowest ramp rate. Resultt of this additional review are summarited below:* All PiR plants licensed with tBC models have temperature ramp rates in the Slow range (clOGC/sec for & period of more than 10 set.onds prior to rupture.)* For the category of plants where t.he POI oiturs dowtritreim of the ruptured nr,^d (it, st'am cooling) the DC Cook. plant is the most sensitive to the NRC rupture/blockage model beculue it hat the tlnwp.t ttmpel-At'WF'

ramp rate prior to rupture.Application of the NRC rupture/blockage model to the plant with the 0iowest tamp tate (DC Cook) Shows that the turrent FNC PttM' MnndI ti. untberattiVt

&s discussed in Reference

(1). thusr, it. i!, euutlutJvd tt;t the currtht tNC tCCS 6nalyses for plants which tall in this cate gory are conservative.

Thest plants are Palisades, Kewaunee and Prairie Isllnd 1 And 2.

T i rrl .Nknu* Par one WH Ah~tyt+/-d 0thltt (At dIihhA) thl P4 I bc~ti A~ hlth uipturad node Ind early in the 1-fil pod p etiod ThA Mf~a I~ h*n ruturf/blockage model tb ft Ginna hat bmm 1akutetdo the tdtrot qodbi§PC? than 6e "at found to be ltes tha" a 206r IthcIs wit 1h uth Pv? still more than 20U6F below the f200tU Ilimtit.& T'he remainingp tiC anhltyild P6" lap.nt. Oh ftabinstrh)

db~5 h8t NAVI&Steam tonling period nor dues thts Fit occur at thP 1-U turod nodi.fo Ir this Plant the' ruIpture Straifl calculdtO~

by the NRC tuptiturs htrkift iftoda (considering the tamp rate ueIfpct oh tuptura. ternperhtuefi)

I getrthatn the- tupture Itrain calculated by th E Ml Thuj, Sto CUptutcq/blockage model would yield is low~er P'CT Siir the "biautf hlq ier Ma~d strain on the hotn-ruptured PC? ttudej Would Impruov& Lid tool in.In guutmary.

it is concluded that Application at~ the ftRt 1uptut&blotkA06 Model in the ENC tCCS model would hot Affect liet~h~itq limits Oil ENC Plants because:-O CT's would be reduced by Using the MrC rtipture/blockagis hodel in all olants in which PCT does not occur an the 1tUPtUred hafde I in thp nite P ant Where Ptl` dofig Occur at' thp 1ruptut~d flod (tp tt the impact of the ftRC rupture/blockwjt'

tunndl On OvC is leA than favl with imtre than A 9006F margin to tha 22Q0tF! limit r?-hn&intg.

Utiton kutil#&r ttnim0hy GENERAL* ELECTRIC GENERAL ELECTRIC COMPANY, 176 CURTNER AVE., SAN JOSE, CALW*FRNA

95125 MC 682, (408) 925-5722 NUCLEAR POWER SYSTEMS DIVISION MyN 278-79 November 16, 1979 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: Reference:

GE CLADDING HOOP STRESS AT PERFORATION

(1) Letter, R. H. Buchholz to D. G. Elsenhut (NRC), ORNL Cladding Swell and Rupture Data -BWR Evaluation, November 2, 1979.(2) Draft Report, R. 0. Meyer and D. A. Powers, Cladding Swelling and Rupture Models for LOCA Analysis, October 31, 1979.(3) General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 1OCFR5O Appendix K, NEBO 20566, January 1976.(4) Letter, A. J. Levine to D. F. Ross (NRC), GE Loss-of-Coolant Accident Model Revisions

-Core Heatup Code CHASTE05, January 27, 1977.During a telephone conversation between GE and R. Denise of the NRC Staff on November 14, 1979, additional information to that transmitted in Reference

1 was requested.

Reference

1 outlined the reasons the data contained in Reference

2 did not affect the GE LOCA cladding swell-ing and rupture models (References

3 and 4). It is GE's understanding that the NRC Staff is concerned with the method used to calculate the ramp rate (clad heatup rate) during a LOCA as it affects cladding hoop stress versus temperature at perforation.

The purpose of this letter is to address these concerns.Section I.B of GE Appendix K Topical Report NEDO-20566 discusses fuel swelling and clad rupture thermal parameters.

this analysis was based on our previously applied CHASTE04 model. Extensive sensitivity studies GENERAL1 ELCICt'iC U. S. Nuclear Regulatory Commission Page 2 were carried out by GE to prepare for NRC review of the currently approved CHASTE05 swelling and rupture model. These studies aro of direct relevance to the current NRC concerns.

The sensitivity studies (results of which are included for completeness in Supplement A) indi-cated only a small sensitivity of PCT to variations in cladding strain and hoop stress at perforation.

In particular, Figure 4 of Supplement A depicts the variatton of the hoop stress at perforation with temperature.

The lower bound of the investigation has been re-plotted on Figure 54 of Reference

2 (attached).

This figure shows that the lower bound of the CHASTE05 sensitivity analysis produces a more conservative relationship of hoop stress to perforation than the 0C/sec curve for temperatures above approximately

740'C (i.e., perforations are not expected in GE BWRs below 925 0 C). The change in PCT for this lower curve compared to the base case was -50F.We understand that the Staff is also concerned about the statistical significance of the range of values over which ramp rates are deter-mined. The calculated cladding heatup rate for GE BWRs is between 1°and 71F/sec. This range of heatup rates is based on an average value over the ballooning portion of the ramp.The foregoing discussion, together with Supplement A, clearly indicates that the ORNL data for hoop stress at perforation for several heatup rates does not impact the conclusions of References

3 and 4 over the BWR ranges of application.

I sincerely hope that this resolves any questions you may have regarding this matter as it pertains to the BWR.Yours truly, R. H. Buchholz, Manager BWR Systoms Licensing Safety and Licensing Operation RHB:cas/4J

Attachments cc: G. G. Sherwood R. Mattson (NRC)R. Denise (NRC)L. S. Gifford (GE-Beth)

Supplement A CHASTE05 SWELLING AND RUPTURE MODEL Sensitivity Studies To evaluate the effects of the change in the calculation of the grey body factors (GBF) in the CHASTEO5 code, a number of sensitivity studies wore done. The studies show that the more realistic calculation of the GBF's results in a smaller sensitivity of the peak cladding temperature (PCT) to various parameters of the rod swelling and rupture model.The studies were performed for a plant with 7x7 fuel at high exposures, to maximize the number of perforations and hence any sensitivity of the calculated PCT. The plant selected had a relatively long reflooding time and a shorter blowdown period which then results in a longer period over which the rods are calculated to be perforated and hence a greater sensitivity to change; in the swelling and rupture model. The results presented here can be considered representative of those expected for BWRs with fuel where perforations are calculated to occur.The following studies were performed and are discussed in detail below: 1. Variation of cladding strain at perforation

2. Variation of swelling initiation criteria 3. Variation of thermal expansion coefficients

4. Variation of perforation stress versus temperature curve 5. Variation of plenum volume 6. Vpriation of the GBF calculation time The base case for all the calculations was calculated using the strains, perforation curve, strain rates and other models as described in NEDW-20566, i.e., nominal strains of 16% on outer rods and 23% on inner rods for perforation hoop stresses <1500 psi. The temperature transients for several rods for this case are shown in Figure 1. Figure 2 shows the relative positions at the different rods.In general, the use of CHASTEDS instead of CHASTE04 results in a smaller sensitivity to changes in various parameters.

The two major reasons for the smaller sensitivity of the results to the changes in the parameters are: a) A more accurate calculation of radiation heat transfer in CHASTE05 has reduced the impact of radiation heat transfer degradation when rods are calculated to perforate.

b) Better nodalization of the cladding in CHASTE05 (it has two cladding nodes instead of one in CHASTE04)

and better control of the time step has reduced the sensitivity of the temperature response to inside metal water reaction as a result of perforations, i.e., when a rod is calculated to perforate, the code takes a small time step.NS: cas: at/4T 1

1.0 Variation of Cladding Strain at Perforation The values of strain after perforation used in the calculation ar based on the FLECHT Zr2 tests described in Section I.B.2.4 of NEDO-20566.

It 15 assumed that for rods with hoop stress (1500 psi, rods next to the channel will have a maximum strain after perforation of 16% of nominal radius and for the reinianng rods, the maximum strain Is assumed to be 23% of nominal radius. The purpose of this study was to determine the change in the temperature response of individual rods and the peak cladding temperature of the bundle as a result of changing the various assumptions regarding the assumed perforation strain. The base case for this study was done using the nominal strains (i.e., 23% an Inner and 16% on outer rods).The study shows that there is a very small (*5F) sensitivity of the PCT to changes in the perforation strain. This Is because, even though individual rod temperatures are affected (by as such as 20 0 F just after a rod perforates during the transient), the temperature of all the rods in-the bundle tend to equalize as a result of redistribution of energy by radiation heat transfer, consequently the overall effect on PCT is small. The studies show that as the strain is increased on an individual rod its temperature decreases, because for larger strains there is a larger area for heat transfer and, hence, lower temperatures.

For smaller strains the temperatures are higher as the area for heat transfer Is smaller.The results for the different cases are presented below;1.The strain on the first rod to perforate (Group 12)was changed to 40%. The calculation showed no change in the PCT but did show a slight decrease (c20F) in the temperature transient for the first rod to perforate shortly after the rod perforated.

Case 2. The strain on the second rod to perforate (Group 10)was changed to 40%. In this case, the PCT decreased by SF compared to the base case. The change was larger compared to Case 1 because of the closer proximity of the final PCT rod to the second rod to perforate;

but despite the change, it should be noted that the change Is small.Case 3. The perforation strain on all rods was set at 30% (30%represents the maximum strain that adjacent rods can expand to without touching).

The PCT decreased by only 3F even though the variation in individual rod temperatures during the transient were lower by as muchas 25F during the transient, just after the rod perforated.

Case 4. The perforation strain on all rods was reduced to 1aTtfie nominal value and the PCT decreased by 3F. In this case also, the individual rod temperature transients changed by a larger value (up to 15F at certain times in the transient)

compared to the PCT.1S: cas/4T 2 The above studies were supplemented by studies using the strains measured in the FLECHT Zr2 test, instead of the nominal strains used in the above studies.Case S Strains measured in the Zr2 test (shown in Figure 2, p 1-175, NEDO-20566)

were input into the CHASTE code, instead of the nominal strains of 16% and 25% on outer and inner rods, respectively.

Figure 3 shows the effect on the first rod to perforate (Group 12)of using the nominal versus measured perforation strains for all the rods. The difference in the temperature transient for individual rods in the two cases is small, and the differences in the calculated PCTs is zero. As discussed earlier, the reason for the small PCT sensitivity is the redistribution of the temperature due to radiation heat transfer and the fact that the PCT rod at the end of the transient is a nonperforated rod. Early in the transient, the PCT rod is often a rod that perforates (as shown In Figure 3).Case 6 This case was similar to Cases 1 and 2. In this case, the strain on the first rod group to perforate was set equal to 16%, 30%, and 40%. For all other rod groups, the Zr2 measured strains were used. The calculated temperature transient for the first rod to perforate is platted in Figure 3. The results show that for higher strains, the cladding temperature is lower shortly after perforation because of a larger heat transfer area, but there is much smaller sensitivity to strain than was calculated in CHASTE04.

The reason for this is that CHASTEOS has finer nodalization of the cladding and hence a more accurate calculation of the surface temperature and the metal water reaction rate. Also, CHASTE05 has better time step control which results in smaller time steps after a calculated perforation.

Because of these two reasons, the cladding temperature does not increase rapidly as a result of inside metal reaction as was observed in CHASTEN4.The calculation for the 40% strain appears to show a larger sensitivity because of a slight delay in the perforation time.But after a few seconds, the temperatures using the various strains all are about the same. The slight difference in the time of perforation for the various cases is a result of the slight differences in the strain rates before the rod perforated.(The strains and strain rates before perforation are a function of the final perforation strain.)The conclusion from this study is that the cladding temperature of perforated rods is relatively insensitive

(<lOF, 15 seconds after perforation)

and the PCT is almost completely insensitive to the perforation strains and, hence, use of the nominal values is appropriate.

NS: cas/4T 3

2.0 Variation of elling Initiation Criteria CHASTE calculates plastic swelling on rods for all temperatures above a certain temperature.

This temperature is nominally set at 200F below the perforation temperature.

Calculations were done assuming that plastic swelling starts OF, 20OF and 400F below the perforation temperature.

The results show that for the case of OF, the PCT decreased by 3F, and for the 400F case the PCT was unchanged relative to the 200F nominal case. The effect on PCT was small (<SF), and the effect on Individual rod temperatures was also small (c20F), and hence it can be concluded that the use of 200F is still appropriate.

3.0 Variation of Thermal Expansion Coefficients This study was done to determine the sensitivity of PCT to uncertainty in the thermal expansion coefficients of the fuel and cladding material.

The changes in the PCT are caused by changes in the gap conductance resulting from changes in the pellet-cladding gap size for different thermal expansion coefficients.

For larger thermal expansion coefficients, the gap size is larger early in the transient resulting in lower removal of stored energy during the blowdown phase of the transient and hence higher PCT. Conversely for smaller thermal expansion coefficients, the PCT is lower than the case using the nominal thermal expansion coefficients.

But in all cases, the sensitivity is relatively small and is documented below for two extreme cases, I.e..' no thermal expansion or contraction and twice the nominal thermal expansion coefficients.

For zero thermal expansion coefficients for both fuel and cladding, the PCT decreased by about 25F; and for twice the nominal expansion coeffi-cients, the PCT increased by about 15F. The small sensitivity to thermal expansion, even for the extreme cases, justifies the use of nominal thermal expansion coefficients.

4.0 Variation of Perforation Stress Versus Temperature Curve The purpose of this study was to determine the effect of changing the perforation stress versus temperature curve from the standard curve used. Two cases were studied, one for which the curve was below all the data, and another for which the curve was above a majority of the data points (see Figure 4). The change in the calculated PCT compared to the base case was S6F for the lower curve, and +2F for the higher curve. For the lower curve, the perforations occurred earlier and at lower temperatures;

hence, the effect of inside metal water reaction was minimized and there was a larger surface area for heat transfer on some rods for a longer period of time. For the higher curve, even though perforations occurred later, they occurred at higher temperatures where inside metal water reaction is higher and hence a higher PCT. The important conclusion from this study is that the sensitivity to time and temperature of perforation has been reduced considerably with the Improved GBF calculation.

NS:cas:at/4T

4

5.0 Variation of Plenum Volume The previous study determined the effect of tim and temperature of perforation on PCT. This study determines the effect of the initial plenum pressure on the PCT. To determine the effect of the initial plenum pressure, the Initial plenum volume was varied from the nominal value by +/-40%. For the increased pressure, the calculated PCT increased by 5F; and for the decreased pressure, it decreased by IF. This study shows that the fuel rod internal pressures have an insignificant effect on the calculated PCT. This is primarily because the effect on the number of perforations is also small, i.e., the number of rods calculated to perforate did not change when the volume was decreased (i.e., increased pressure)

but decreased by only two rods compared to the base case when the volume was increased (i.e., lower pressures).

In this study also, the time of perforation changed for the different conditions but the sensitivity was small.6.0 Variation of Grey Body Factor Calculation Time Different rods are calculated to perforate at different times during the heatup transient.

As most of the rod swelling occurs a few seconds before the rod is calculated to perforate, it is appro-priate to calculate GBFs at the time of perforation of each rod. A study was done to show the sensitivity of the PCT to two bounding assumptions about the calculation of G6Bs. In one case, for the radiation calculation only, It was assumed that the perforated rods did not swell. In the second case, it was assumed that at the first perforation all the rods had swollen to 23% strain in the calculation of GBFs.* The second case is the procedure that is used in CHASTE04.

For the case in which no swelling was assumed, the PCT was lower by about 20F compared to the standard case where GBFs are calculated at each perforation.

For the second case where GBFs were calculated assuming all rods swollen, the calculated PCT was about 11OF higher compared to the standard case. Figure 4 shows the variation of PCT for the three cases. This study shows that the PCT using the old (CHASTE04)

procedure was extremely conserva-tive, and the degradation in radiation heat transfer is not very large as a result of the perforation of a few rods, which is the typical case in BWR loss-of-coolant accident calculations.

KS: cas/4T 6

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ROD ThNPERATUES

USING K014INAL STRAIN VALUES I

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2500 qtax eooo/5o00'400*500 TR*41T1014G

UNCOVERY)1

0 P. 4 to FIGURE 5 EFFECT OF CHANGING GBF CAL CULATION PROCEDURE emoeplew1^01--el aA 046'* AL ,&OyAwhom

-,,O r^ls -ro" esrhfcv,4.

s. -SPWRAY START IRST P MOProI~rON

OCCURS 0 100 aojof'%2oW

z I GE &BURST TEMVIPERATURE

CURVES f I I: 5 la 15 ENGINEERING

HOOP STRESS 20 (KPSI)FIGURE 54 OF REFERVNCF

2 Westinghouse Water Reactor NLM3r TechrKcay M.vit'.n Electric Corporation Divisions Box 355 PitTsurgh Penrnii- Wama U November 16, 1979 NS-TMA-21

63 Mr. Darrell G. Eisenhut Director, Division of Operating Reactors Nuclear Regulatory Commission

7920 Norfolk Avenue Bethesda, Maryland 20014 Dear Mr. Eisenhut: Letter NS-TMA-2147, dated November 2, 1979, responded to NRC concerns related to the fuel rod models used in the Westinghouse LOCAIECCS evaluation model and potential non-compliance with the requirements of lOCFRPart

50. Table 1 of that letter included information on fuel rod heatup rate prior to burst. That information was based on our initial evaluation of the results of current LOCA analyses for Westinghouse plants with operating licenses.

Subsequent to completion and transmittal of that letter,.Westinghouse continued investigation of heatly rote calculations.

As a result of that investigation, Westinghouse then developed a procedure to determine clad heatup rate prior to burst. That procedure keys on the calculated clad strain during the LOCA transient to establish a starting point, in time, to use in the heatup rate-calculation.

That procedure was presented to NRC personnel during a meeting on November 13, 1979, in Bethesda, and was accepted on an interim basis, as adequate with respect to Appendix K LOCA analyses.

Table A shows the revision to the heatup rates previously given in Table 1 of Letter NS-TMA-2147.

Inspection of Table A shows heatup rates, in some cases, less than 250F/sec.In the current W ECCS Evaluation Model (Feb, '78) used for the above analyses, a fuel rod burst curve which represents burst conditions for heatup rates of 25 0 F/sec and larger was used. From Table A, since some cases have heatup rates less than 250F/sec and burst conditions change for lower heatup rates, Westinghouse recognized that some of those analyses could be non-conservative with respect to the time of rod burst.Therefore, W performed an evaluation of all operating plants licensed with the W ECCS Evaluation Model with respect to use of a heatup rate dependent bursT model. The heatup rate dependent burst model currently used in the W Small Break Evaluation Model (documented in WCAP-8970-P-A "Westinghouse Emergency Core Cooling System Small Break, October 1975 Model" and approved by the NRC) was used in this evaluation.

-NS-TMA-21f3 November 16, 1979 Page Two The results of that evaluation, the status of each plant evaluated and justification of conclusions reached are as follows: PLANT (1) MODEL FEB. '78 FQ 2.31 PCT 2172 A new analysis was performed using the appropriate heatup rate burst curve and water residing in the accumulator lines (not previously accounted for) was considered.

The resulting PCT was 2135 0 F at an FQ of 2.31.Therefore, lOCFRSO criteria are satisfied.

PLANT (2) MODEL OCT. '75 FQ 2.17 PCT 2199 A LOCTA run was made using the Oct. 75 evaluation model with appropriate heatup rate burst curves for FQ 2.16. PCT -2127 Use of Feb. '78 evaluation model, in particular the new accumulator discharge model, will compensate for the AFQ, shown above, to maintain 22009F. (This is a burst node limited plant)PLANTS (3) (4) (5) (6)Since the heatup rate for the hot rod is greater than 25 0 F/second and the PCT does not occur during the steam cooling period, the current analysis for these plants remains valid.PLANT (8) MODEL OCT. '75 F 2.10 PA 2188 F An Oct. '75 model LOCTA run was made using appropriate heatup rate burst curves. Results were: FQ -2.10, PCT -2227.Application of the "Dynamic Steam Cooling" modification of the Feb. '78 evaluation model will result in a 60 0 F reduction in PCT and the Feb. '78 accumulator discharge model will result in at least a 20 0 F reduction in PCT. Results of a Feb. '78 model analysis are expected to result in a PCT of approximately

2147 0 F at an FQ of 2.10.Therefore, lOCFR5O criteria will be satisfied and there is no safety concern.PLANT (9) MODEL OCT. '7'FQ 2.25 PCT 2142 Novemni1er

12, 1 979 Page Three Based on the results of a calculation for plant 1(14). the use of approximate heatup rate burst curves would result in a maximum PCT increase nf 680F. Thus, the estimated (maximum)

PCT = 2142 + 68 = 2210 0 F at an Fq -2.25.The benefits associated with the Feb. 78 accumulator discharge model and accounting for paint on containment heat sinks will result in a PCT reduction well in excess of lOF.Therefore, no safety problem exists.PLANT (11) MODEL FEB. '78 F 1.90 POT 2124 A LOCTA calculation was performed using appropriate heatup rate burst curves. An F of 1.89 resulted in a PCT of 21610F.Therefore, a peaking factor reduction of less than 0.01 is required for this plant to remain in compliance with lOCFRSO.PLANT (12) MODEL OCT. '75 Fn 2.21 P1,T 2198 Based on analyses performed for plant f(15), a 15F/second reduction in clad heatup rate impacts hot rod burst to effect PCT by +42°F. Extrapolating, a 170F/second reduction in heatuD rate results in a 480F PCT increase.

Use of the dynamic steam cooling calculation on the accumulator discharge model in the Feb. '78 ECCS evaluation model results in an estimated

(60 0 F + 20 0 F)80OF reduction in PCT.Therefore, a Feb. '78 model analysis would result in a PCT of 2198+48-8O=2166 F at F of 2.21 and no safety problem exists.PLANT (13) MODEL FEB. '78 Fn 2.05 P12T 2172 A LOCTA calculation was done using appropriate heatup rate burst curves and the results were: F -2.05, PCT 2191F Q Therefore, no safety problem exists.PLANT (14) MODEL FEB. '78 Fn 2.32 PET 2124 NS-1MA-2163 November 16, 1979 Page Four A LOCTA calculation was done using appropriate heatup rate burst curves and the results were: F -2.32, PCT -2192OF Q Therefore, no safety problem exists.PLANT (15)MODEL FT PET FEB. '78 2.32 2158 A LOCTA analysis was done using.appropriate heatuo rate burst curves and the results were: FQ -2.32, PCT = 2200OF Therefore, no safety problem exists.PLANTS (16) and (17)The latest licensing analyses have been verified to use appropriate heatup rate burst curves and therefore remain valid.PLANTS (18) and (19)New LOCTA analyses.

were performed The PCT was virtually unchanged.

using aoDrOpriate heatup race burst curves.Therefore, no safety problem exists.Based on the detailed information provided above, the Westinghouse Safety Review Committee concluded that two plants were found to require a reduction of 0.01 in allowable core peaking factor to maintain a PCT of 22000F. Four other plants have current analyses to the October, 1975 version of the Westinghouse model and may require a peaking factor reduction.

However, we.believe that reanalyses with the most current Westinghouse LOCA/ECCS evaluation model (February, 1978) would show that no changes are necessary.

That is, we believe margins available in this model will more than offset any effect associated with the change in the fuel clad burst curve. A copy of the NRC notification letter (NS-TMA-2158)

retarding this iss;;o is attdcheu.The above information was also presented to the NRC Staff at the November 13, 1979 meeting.Following the November 1, 1979 meeting, Westinghouse has again reviewed the, ORUL data quoted as a basis for NRC concern regarding adequacy of the W Appendix K blockage model. Comparison of individual rod burst strains from ORNr data to the corresponding Westinghouse data which has used as a basis for our blockage model indicates the ORUL data is in excellent agreement with the W data. Since the axial distribution of the burst strains in the ORNL multi rod '3urst test has I-NS-TMA-2163 November 16. 1979 Page Five been shown by ORNL to conform to local temperature distributions in the specific heating rods used in the tests, conclusion as to the applicability of the axial distribution of bursts (which is the Darameter that relates individual burst strain to flow blockage)

cannot validly be made. Never-theless, the blockages measured from the ORAL tests are similar to those calculated by the Westinghouse model, which has been approved by NRC, when due consideration is made in translating blockages measured in 4X4 bundles to blockages applicable to 15X15 or 17X17 rod fuel assemblies using accented statistical techniques.

Thus, we believe no immediate action is aoDropriatp.

with respect to reanalysis of Diants using the proposed NRC blockage model pending detailed review of the proposed model.As a result of further investigation and evaluation, the following can be concluded:

1) A modification to the W model to account for the heatup rate dependence is necessary for compliance to Appendix K.2) The impact of this modification is relatively small, effecting only two ooerating plants in terms of requiring peaking factor adjustments to meet the criteria of lOCFR50.46.

The affected utilities ana the NkC ndve been czdeiqadLly inifurr-ied.

3) Comparison of the Westinghouse data and ORNL data shows excellent agree-ment and the current Westinghouse model, in the range of interest, is still appropriate.

It is therefore concluded that no safety problem for Westinghouse plants has been identified and all plants are in conformance with NRC regulations since the burst temperature modifications

(1 and 2 above) are accounted for.Very truly yuurs, T. 14. Anderson, Manager Nuclear Safety Department NS-TMA-2163 November 16, 1979 Page Six TABLE A REVISION TO HEATUP RATES TRANSMITTED

IN lETTER NS-TMA-2147 CASE HIEATUP RATE (0 F/SEC)HOT ROD AUG OR AW ROD 1) 8.5 10.9 2) 20.3 13.1 3) 25.6 18.0 4) 25.0 15.4 5) 31.5 19.4 6) 27.4 23.8 7) (Not Westinghouse Fuel)8) 19.1 7.4 9) 12.3 12.0 10) (Not Westinghouse Fuel)11) 6.2 11.3 12) 8.0 11.4 13) 18.3 16.1 14) 9.3 14.3 15) 8.2 13.8 16) 39.6 23.7 17) 43.2 26.7 18) 22.7 17.6 19) 26.5 16.7 Westinghouse Waler Reaclor PBtothPxI) ,AL Electric Corporation DiviSionS November 16, 1979 NS-TMA- 2158 Mr. Victor Stello Director, Office of Inspection and Enforcement U.S. Nuclear Regulzatory Coornnission East West Towers Building 4350 East West Highway Bethesda, MtD 20014 Dear Mr. Stello: Subject: ECCS Evaluation Model This is to confirm our telephone conversation with Mr. Frank Nolan on Friday afternoon, Noverrmer

2, 1979. In that Conversation we reported a non-conserva- tive feature in Westinghouse large break ECCS'evaluation models.The Nuclear Regulatcry Com-mmission staff met November 1, 19iW, with representa- tives of reactor vendors and nuclear fuel suppliers

-- Combustion Engineering Inc., Exxon Corporation, General Electric Company, Westinghouse Electric Corporation and Babcock and Wilcox Company. Utilities which operate nuclear power plants were informed by NRC.The purpose of the meeting was to discuss the staff's ongoing evaluation of the results of tests on electrically-heated fuel assemblies conducted at the Oak Ridge (Tennessee)

National Laboratory, .JRC indicated that emergency core cooling system analytical codes currently used to evaluate the effects of postulated loss-of-coolant accidents (LOCA) might not be in compliance witn NRC regulations.

The portion of the codes in question deal with the effects of fuel clad swelling and rupture and blockage of cooling water.Subsequent to the meeting, Westinghouse performed a detailed evaluation of the most recent analyses for operating plants and on November 2, 1979, Westinghouse confirmed, in writing, that the impact of the information pre-sented by the NRC has negligible impact on the LOCA analysis results of tne plants licensed with the Westinghouse LOCAIECCS

evaluation model. The %RC staff has concurred with this conclusion.

<-Mr. Victor Stello -2-. NS-TMA-2158 However, as a result of that detailed evaluation, Westinghouse has now recog-nized that .non-conservative feature could exist In-the Appendix K LOCA analysis with respect to the portion ot the calculation reltited to tue' rod burst. The potential non-conservative feature of Westinghouse larget break ECCS evaluation inodels is as follows. The models use a curve which represents fuel clad burst conditions tor clad heatup rates of 25F/second and yreater The evaluation discussed revealed thdt heatup rates could he less than 25*F/second.

During the LOCA tra~nsient, thle tuel cldd burst curve dstablishes the time of clad burst 3nd (since :le clad temperature ana the pressure dit-ferential across the clad 3re.changing throughout the LOCA trenslert) "ne post-burst conditions of the clad. The fuel clad burst curve is deoendeft on the clad heatup rate prior to burst and a reduction in heat.Ap rate ca.stes earlier clad burst. A shift in clad burst time can affect the peak clac tem-perature (PCT) calculated for the LOCA transient.

Therefore, in order to more fully evaluate'this effect, the clad heatup rate prior to burst was determined from the most recent LOCA analyses for ti'osd plants licensed with the Westinghouse LOCA/ECCS

evtluation model. Plants having heatUp rates less than 25 /seccnd were reanalysed to ascertain tne effect on peak clad temperature.

Two plants (Turkey Point 'Jnits 3 and d) were found to require a reduction of 0.01 in Fg to maintain a peak Clao Tem-pera-ture (PCT) of 2200 F. A third plant, Indian Point Uni' ac. 2, was nct expected to reculir any FQ recuction, considering -he prsoit PC7. Ind avaii-able sensitivity studies. Analyses, underway at the time of cur telephone conversation, have now been completed and confirm this.Four other plants, currently not operating (Trojan, forth Annd Unit 1, Indian Point Unit 3 -and D. C. Cook Unit 2) have current analyses Lo the October 3975 Westinghouse rrodel and on that basis might require a reduction in Fn. How-ever, we believe that reanalyses dtth the most recently aporoved Westinqho;se LOCA/ICCS

evaluation model (February

1973) would show that no changes are necessary.

That is, we believe margins available in this model will more than offset any effect associated with the change in the fuel clad burst curve.We have advised the affected utilities of this unreviewed safety question.

As part of this overall evaluation, we are examining plants under construction and will report as dppropriate.

Please teel free to contact Dr. Vincent Esposito (412-373-4059)

if yuu should have any questions.

Very truly yours, T. M. Anderson, Manager Nuclear Safety Department

/wpc I. 0; ^T:Vwnr: .IttieS4 -4C PAE -I 3 YANTEE ATOMIC ELECTRIC COMPANY fiG mSr-3*Q,-07fl niteS Stas Ese- kgitory nsss taSX e >, 9 e Attntiat t~efer-Of f f le tor Xteulatitm I-f. Darell tizezstkt ketin'r-Iraw

(1) Lit N. P71-3 (Docket No. .1C-i (2) Ietter ta lEfl Mt "Tvalmti tf Ce dn sl ai4'4 Ewtv~n Nodels, t dated Nontber 2, IS79.(DUIf NaG. 0630 dAted 111S179, entitfle, "Claddin trilling cn- tuptrs Pa-Ilc foar L= Anilywit.(4) fif~4, 'Lnn Nule~r o~cy~-kuu aGenric tM ra ti Nd Upt Etv--11, ily 1976a De=r Sir:-Snh-jfl -Snstia of CleAdin, Swlling -w el This letter is -aAdakn4 to : Fsulitted- toyn an trmlcr 2, l97l, Lterrnne t. U is ftrrizc in reponse to DitiWAf £Vweicmu raises VI your staff cocrnng te haling of cl*ding bc-irp rate depentnCe iu T's li C- dels fw clin 1 afliug mae-rp.. The Ifoi am hopefully respauive to YOE qwstitx¢ £t tk sltle t te rt cktr orretia lt irri& taperutc regim n 000e ( 10C or grectorY Tfhis Is X t the reltively

1w f11 pa rpmwure is ]thI Take twA IU. IC thisr-t taseratn range, rn'ttr tesw ttwv is oxtresely sauitiuw to o -CflcumtJrap twera ftr E taui- Zf jt Writ a 1w,, a a t 543. C/5OCsa.o At rresat, Tain nes strz Ccrrwitic, Wcdified for Yats RCm 4wintry, -of Wrst taperature vs.. I Taint's Wrst strain n-ttrufl- aru9titi6Ci averpredicts th- Stais capred t ~ hs cwrrulstzca n%- ct rap~-rat&

e £sjtret. With regard to rlnt

'.04 -KY -1. 5. Thtlnr Zepl stw Cciac Uo!g 70 19 tnperature, we currently perceive tbe corretctin of burnt ttr-xturw-v*, %tres to 'be ap1inb1. to afl ,rmp ntn ic high teprattre ruptnre. With regard to bar-it atroiu, tek &rrelations that envelope the slcwrnp ani fast-rasp fLC draft ur-£retatha of bunt train vs. turmr4t tay -o y not be cors tin eit n#wtlwr cmt tAlcutste clad ruptua.This pvscptIcm of rp rate imsensitivity at high tflptrtztms way be modifitd as -Mr Ut-s in th& lrNt ttatup rY TS, hi_% biWrw tmIqxprtttfl ngicn is assessed.

There apears to be a ratstepcmec AA2octiAted with Icktr tetperatce burst. no clear Jstifieation existe s f extrspolati of.w rarw teperate rzqtnre data to hit tee(veratutrt rtpture eictiticns., On czia handzlple rod brtdata frIn nxeriats dons ina tn wA r to be limted to high (2t 0 Chia.)tww"rltre r=pS in this high te ertur ngi. 0. the other , limted dnta frcu sh&Is rod tests in wems, for ezaple, indicate tazt ra dspen4uuve at bigb t*tperatnra coul1 be possible.tan'" has Ammd tM i pct of atilizit tbe fC draft turns.s"dasted with sio rp rate effects (dfetrca 3) by per ca nw-2 cal~cluticn vrwerk£ tin exposure Jistributica for th reainder ofb te prn"vt cycle. Than calcialatians ban teen pqrtormeA in the fcflcmTo amr: CD) Mi- S= 0%C/se-cuen for burnt tepr In. *trus was mified to reflect T ' rgtu of ba3yJse. by iterrati latwn= thu OtCnc. ad -te 2VC/aa. rrelatitn touaed in refi 0) e ai-ras p burst ain- (7igure & of reference

3) was used in tW~-2 satbouj* n data points grs *usbsin the- hi (3) thw sir-rnm iocal fiw biocka" c- ofvnftc e 3 was med.a;g with the I'-WXfl-j (refereoce

4) 11w tt umdttplin tunes associted with 20% rdrteicu La fleot aresX, oeistat with the prutdidn of the time-rap local blockap *cc-.sed an this aysis, Tn* c ed tt r prt.the pleat for tes raaindut of tb* pryet cle Is zinuffecced by the Irluzsia of the attn r rate draft corralatis in the TIO liSed 1.d A rmeuti in i sd Titr r w d cated mnmy. Spe ufil fAr frIh tsl a i. present expo l r bea-titw Of 9.60? wIlft meets Appendix 4 critais with the 7CV-Z ucxifieatita mted shov.; 9.550 kw/ft is t l pner oPeration..

This fuel is CUrrently licese St1d tm7/ft At tm present~Vezm

• T the MiO pr etposA fuel, Pf bufit Apendix I nd 9.4 r/ft it neefk for full wr ft b-CrrZeatly limied. Etd of cycle oO eaft Irm thee i l id the Cmis .-. l to etasize that jcu: K. -r* 4.V. S. tla ,ttr~terv ot;*f ic Lw-r e 20 1979 we d nr -CZsZ1scy euppprt tim nlidty Of the c tin rsfccre S cA that ! p1 it Et. od sky krate ke -- s opraE n the bc4nain of Cj1 s~taO! 1 nfp hxbe avaafl to Q& tse anlyes If there it little beutuprqate dee e ot rupture t t for hig ttmpta-tv.i- failure, tes enld gl4 q6=f t t=-= posil inpaCt of low te-ea--tare rate tube ruptar fl tYak -ee M cppiia-eaittcas by t all trh Orto/nc. 'tx jz wXIT " I s lbs f-0 C s data fr ntuTre terrratnx fl10C. A quadrmtic fit thrt@- this &t* Fet yialdz tOE foli-win Vorelaticn:

t T' F a "7.75 -IMQ p. o.ooz e42 (2i DAI; ?I in &6grn , AT inzi)Substitntf!

this c1prrtla1tio i ntt 7VQf-, Iz ith the c 'c strain corz-elsth;

of n n 3 c fta tbht th_ peat ct.!teaperatudre for fresh fiel at £,O PMr= (the most t ti it Lim exposw-c rcgic caawidred abowF) is JJT0 le thE Ycket'r airreomt lfeiee luedstel predicts.it"se c idtratita Yu tet the fol (1) Knp depenent c- sr asratlited with ca-dng swelling ski rupture nAela shnold ld t affet plat e tit for t (2) j-aze 4 rn= r wruaticu of tht dat of reference

3, (and in prart c te paucity of data appropriate to cleddin-4 swellng Amd rutr rrfw at Toalx Rwr) w.e cwwwitvt tbat IUm 's arnt ( ) CoottAilo arqttCiritim ad evAlna-ticn of i LUs, pitulty i th $-c raW ratc 1 Lit lPeraturC

Ingiotii requirred.

During IP, till k* ewroacbing t flZ with a. rcz4 kntupm odal to re-place*

U -2As pest of ou Uenr EC- ca"S. This m.odel, wiln MdTe ni ATable data.(4) ParricuIla attenle stould be placed c' the strain a btzrst data.amd ft nalaricnbh to rwrtrt ivt cla ;resure zan £-d If yo 1 Ay quenties resadixn this letter, please fuel Ie to concz Dr. LAss!f mac or Dr. Stephen?.

tehpitt of rItle tnu iUnri rizg ,R r--tQfw vCvwf-z Sire i7.Mmy Twm &ImC izceIC Wwr U. j4&m tA. 4a£Cflna w.Mr. William J. Cahill, Jr.Consolidated Edison Company of New cc: White Plains Public Library 100 Martine Avenue White Plains, New York 10601 Joseph D. Block, Esquire Executive Vice President Admi nistrative Consolidated Edison Company of New York, Inc.4 Irving Place New York, New York 10003 York, Inc.Ms. Ellyn Weiss Sheldon, Harmon and Weiss 1725 I Street, N.W.Suite 506 Washington, D. C. 20006 Joyce P. Davis, Esquire Law Department Consolidated Edison Company of New York, Inc.4 Irving Place New York, New York 10003 Richard Remshaw Nuclear Licensing Engineer Consolidated Edison Company of New York, Inc.4 Irving Place New York, New York 10003 Anthony Z. Roisman Natural Resources Defense Council 917 15th Street, N.W.Washington, D. C. 20005 Dr. Lawrence R. Quarles Apartment

51 Kendal at Longwood Kennett Square, Pennsylvania

19348 Theodore A. Rebelowski U. S. Nuclear Regulatory Commission P. 0. Box 38 Buchanan, New York 10511 John D. O'Toole Assistant Vice President Consolidated Edison Company of New York, Inc.4 Irving Place New York, New York 10003