ML17249A715

Forwards Info to Support 791220 Application to Load Four Mixed Oxide Fuel Assemblies in Cycle 10,in Response to NRC 800213 Request.Westinghouse Ltr Re Proprietary Affidavits Encl.Response to Question 6 Withheld (Ref 10CFR.790)

ML17249A715
Person / Time
Site:	Ginna
Issue date:	02/20/1980
From:	White L ROCHESTER GAS & ELECTRIC CORP.
To:	Ziemann D Office of Nuclear Reactor Regulation
Shared Package
ML17249A716	List: ML17249A715 ML17249A717
References
NUDOCS 8002250526
Download: ML17249A715 (35)

Text

REGULATORY i ORMATION DISTRIBUTION SYS M (RIDS)

ACCESSION NBR:80'02250526 DOC ~ DATE: 80/02/20 NOTARIZED: NO DOCKET FACIL:50 240 Robert Kmmet Ginna Nuclear Planti AUTHOR AFFILIATION Unit ii Rochester G 05000240 AOTH ~ NAME NHITKe L ~ D ~ RocFiester Gas 8 'Electric Corp, RECIPBNAME RECIPIENT AFFILIATION Z IEHANNg D ~ L Operating Reactor s Branch 2

SUBJECT:

Forwards info to support 791220 application to load four mixed oxide fuel assemblies in Cycle 10ijn response to NRC 800213 request.westinghouse ltr fe proprietary affidavits encl'esponse to Question 6 withheld (ref 10CFR~790) 3P ~

DISTRIBUTION CODE: PAOIS COPIES RECEIVED:LTR ~ ENCL /+BAIZE:

TITLE: Proprietary Info after Issuance of License

~O~~S: LQ ' ggA4'tbSSa, 4. Mue~4gAM RECIPIENT COPIES REC IP IKNT COPIES ID CODE/NAME LTTR ENCL ID CODE/RANE LTTR ENCL ACTION; .02 BC Og g 4/~ 7" INTERNAL: 0 RKG FILE 09 I 8 E 2 ~fat@'6 11 1 2. 12 CORK PERF BR 14 ENGR BR 1 1,93 15 REAC SFTY BR /4 16 PLANT SYS BR 17 EEB e- nit 1

18 EFFL TRT SYS 1 NRC PDR OELD 1 0 STS GROUP LEADR 0 Ql8'+A S8 EXTERNALS 19 ACRS 16 16 LPDR NSIC 1 4 AP SMITHEH ~ H/AFF 0 EES )E 7 9 ( > qSOit'i

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NUMBER OF COPIES REQUIRED: LTTR ~ ENCL

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DOCiZT NO. SQ DATE: p 2 CP 8+

NOTE TO NRC AND/OR LOCAL PUBLIC DOCUiiENT ROOi~iS The following item suomitted with let"er dated Z. - Z,O P~

from ac er 8 E/e,e4 i'c, is being withheld from public disclosure in accordance with Section 2.790.

?ROPRIETARY INFOR:~iATION ide ual Distribution Service's Branch

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)' IIA11 ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.Y. 14649 I.EON D. WHITE. JR. TKI.KPHONK VICK PRKSIOKNT ARKA COOK TIK 546-2700 February 20, 1980 Director of Nuclear Reactor Regulation Attention: Mr. Dennis L. Ziemann, Chief Operating Reactors Branch No. 2 U.S. Nuclear Regulatory Commission Washington, DC 20555

Dear Mr. Ziemann:

By letter dated February 13, 1980, you requested additional information to support our application of December 20, 1979 to load four mixed oxide fuel assemblies in Ginna Cycle 10. Attachment A to this letter provides our responses (non-proprietary) to your questions. The response to question 6, which addresses densi-fication of Ginna mixed oxide fuel, is deemed to be proprietary by its owner, Westinghouse Electric Corporation. The proprietary response is provided in Attachment B to this letter.

As this submittal contains information proprietary to Westinghouse Electric Corporation, it is supported by previously submitted affidavits signed by Westinghouse, the owner of the information. The affidavits set forth the basis on which the information may be withheld from public disclosure by the Commission and address with specificity the considerations listed in paragraph (b)(4) of Section 2.790 of the Commission's regulations.

Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10CFR Section 2.790 of the Commission's regulations. Correspondence with respect to the proprietary aspects of the application for withholding or the supporting Westinghouse affidavits, should reference CAW-80-08, and should be addressed to R. A. Wiesemann, Manager, Regulatory and Legislative Affairs, Westinghouse Electric Corporation, P.O. Box 355, Pittsburgh, Pennsylvania 15230.

A total of forty copies of this letter and of Attachment A and a total of 3 copies of Attachment B are provided. <~+ ~rp Very truly yours, "o ~

P p Jr.

L. D. White, LDW:np 8() pap 505M

Attachment A Responses to Questions on Mixed Oxide Fuel (Non-Proprietary) has a lower melting point than Uo . Was this included in the oNrpower aT <rip equation? If nor, lease explain.

R~es onse The results of the evaluation of transients with mixed oxide (MOX) fuel were shown to be nearly equivalent to previous results (see XN-NF-79-103). As a consequence, the adequacy of the overpower hT trip logic was confirmed for the proposed loading of the mixed oxide fuel.

The mixed oxide bundles are specifically placed in low power regions of the core.. The bundle average power of the mixed oxide bundles are to be well below the bundle average power of the core; hence, it, is anticipated that the mixed oxide bundles will not approach conditions of centerline melt during reactor operation.

UESTION 2 Provide curve of Pu isotopic content and U isotopic content as a

a function of burnup of the MOX and 'UO fuel. Also, the relative energy produced by each isotope for both if available, cases.

R~es onse Analyses have been performed for the mixed oxide fuel assemblies and for the fresh uranium (3.45 w/o) fuel assemblies being loaded in Cycle 10.

Figures 1 and 2 represent the Pu and U isotopic content for the MOX and UO assemblies, respectively, as a function of burnup.

Figures 3 And 4 represent an estimate of the relative energy produced by each isotope for the above cases.

UESTION 3 In Table 1.1 of XN-NF-77-40,'Rev. 1, explain why the minimum DNBR does not change even though the maximum power level and maximum core heat, flux have changed for the steam line break.

~Res onse The analyses of interest involve the large steam line break incident which has been evaluated in both Rev. 0 and Rev. 1 of XN-NF-77-40. The reevaluation of the incident in the Revision 1 version was motivated by a revised estimate of the Cycle 9 pressure reactivity feedback and boron worth in comparison to those values determined for Cycle 8.

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The MDNBR for each case discussed above was calculated consistent with the methodology used for other transient evaluations, and the results are given below:

Rev. 0 Rev. 1 MDNBR 2.21 1.58 These results are consistent with the calculated increase in heat flux for the Revision 1 case, and each value is far above the minimum safety limit of 1.30 for the W-3 correlation.

A reevaluation of the boron reactivity worth during the small steam break transient (as reported in Rev. 1 to XN-NF-77-40) was performed as a result of a review. of the reactivity transient beyond 200 seconds.

This independent reevaluation indicates an underestimation of the boron reactivity worth injected into the RCS by the referenced report. The proper calculation of the total reactivity prediction eliminates the misleading indication that, return to power conditions could be anticipated. The reevaluation was an independent calcula-tion relying only on the RCS pressure prediction from the reported analysis.

A complete reanalysis of the small steam line break incident would be expected to have comparable overall results with only minor variations caused by differences in the more detailed time integration performed in the PTS analysis. In either case, return to power is not anticipated for the steam line break transient, and the original time scale of the reported analysis is sufficient to establish reactor system protection.

Assumptions used in the independent calculation are summarized in Table 1. Table 2 shows the reported boron reactivity and total reactivity worths compared with values reflecting the result of the reevaluation.

UESTION 4 Provide some verification of the ability of the Exxon physics methods to accurately predict. core behavior in cores containing MOX enrichment in the range of those to be used in Ginna for Cycle 10.

R~es onse A discussion of ENC's ability to accurately predict core character-istics of MOX bearing critical experiments is provided in Section 4.2 ("Verification by Critical Experiment Comparisons" ) of XN-75-27

("Exxon Nuclear Neutronic Design Methods for Pressurized Water Reactors", June 1975). In addition Exxon Nuclear has irradiation experience with MOX fuel in two operating BWR's. Forty-two (42) assemblies have been irradiated in Big Rock Point (peak exposure assembly at 30,400 MWD/MT) and eighteen (18) in the Kahl reactor.

TABLE 1 BOROH WORTH'REEVALUATIOH ASSUHPTIOHS Parameter Yariabl e. Value Source RCS Volume (ft ) 5973.2 Boric Acid Tank Reserve (gal.) 2000 (total for 2 tanks)

Safety Injection Lines Volume (ft ) 42 (total )

27.7 (to cold legs only,'1)

RCS Pressure (psia)

Safety Injection Pump Flow (1 pump) (2) ~

No. of SI Pumps Working '2 Boron Concentration in Tanks (ppm) 20000

Reactivity Worth of Boron (hp/ppm) .872 x 10 Temperature of Boron ( 'F) 145 Pump Startup Delay (seconds) ~

10 (1) XH-flF-77-40, Rev. 1, Figure 3.36 (2) R. E. Ginna Technical Specifications (Change Request), dated September 1976, Figure 2.1

TABLE 2 REACTIVITY TABULTIONS BORON. REACTIVITY WORTH ( $ ) TOTAL REACTIVITY WORTH ($ )

~dtd' 1 td(3) ~dd (2) R 1td (3',

0 0 0 0 . 0 180 -. 04 - .51 .79 - 1.26 I

195 -.70 -1.31 -1.28 - 1.89 210 -.74 -2.07 -1.14 2.47 225 -.76 -2.92 -1.00 3.16 240 -.78 -3.81 .86 3.89 255 -.80 -4.71 o72 - 4.63 270 -.82 -5.64 .59 5.41 205 '-. 83 -6.58 .47 6.22 300 -.84 -7.52(") .35 7.03 (1) Frcm ttme of start. of st.earn 1 tne break.

(2) Xf(-l>F-77-40, Rev. 1.

(3) See Table .1 .

(4) Approxtmately hal f of the bortc acid remains $ n the 1njectton l tnes or tanks.

No unusual difficulties have been encountered in predicting the behavior of these cores with the presence of the MOX fuel. It is again pointed out that the Ginna application of 4 of 121 assemblies or only about 3% of the core will have a minimal impact on the overall core characteristics and behavior throughout, the cycle.

UESTION 5 The proposed change to the Technical Specifications for Ginna says that the enrichment of reload fuel is limited to 3.5% of U235 or its equivalent in terms of reactivity.

The equivalence should be specified exactly (a brief discussion=

in the Basis would be acceptable). Under what. conditions is the equivalence to be achieved?

~Res ense The Technical Specification provides an upper limit on reload fuel enrichment. We suggest that equivalence be defined in terms of reactivity in storage in the spent fuel pool. Thus, a paren-thetical phrase added to Technical Specification 5.3.l.c would provide the necessary definition: "(as defined by the spent fuel pool storage evaluation)". For the four mixed oxide fuel assemblies, the spent fuel pool evaluation provided in our December 20, 1979 submittal demonstrates that. the proposed limits are met.

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6. The staff SER on Mixed Oxide Fuel states that use of U02 densification models for tlOX should be verified.{l) Since it appears that this assumption was used for the ttOX fuel for Ginna, please verify this assumption. Does the predicted amount of densification satisfy Reg. Guide 1.126.

The current Westinghouse UO> fuel densification model was used to predict

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the densification performance of the HOx fuels fabricated for Ginna. This empirical model, which w'as derived using only UO> performance data, uses fuel sintering temperature and sintered density to predict the extent of densification. The application of the model to It0 fuel is justified on X

two bases; the similarity in procedure used in fabrication of HO fuel to U02 procedures and the ability to conservatively predict the performance of previously irradiated t~iO fuels.

X A

While.studying the densification of pure U02, the relative dimensional

.stability of fuel was found to be strongly depergent on the pelletizing

.process employed and the values of key processing parameters. The processes used in pellet fabricatio'n for Ginna are essentially identical with those used in preparing HO fuel previously irradiated'y Westinghouse whose X

performance is conservatively predicted by the U02 model. Further, the

.pelletizing processes used for preparing the Ginna tl0 fuel is very similar X

to that used by Westinghouse to produce U02 fuel. The limiting processing

'parameters, sintering temperature and final pellet density, were controlled to a range which produces UO> fuel with. predicted performance within design requirements.

The Westinghouse experience with MO densification performance during X

irradiation has been reported and compared with model predictions in WCAP-8349-P. Theoretically, HO densification would be expected to proceed at a lower rate than in UO> since pore removal is directly related to local fission events. Densification or pore removal is a bulk process and in U02 fuel the fission events are fairly uniformly distributed in the material and the product densifies uniformly.

. However, in hOX the fission events are concentrated in or very near to Pu0> particles which are free to densify independently of the UOp matrix. The low enrichment UO~ matrix experiences a lower fr'equency

. of fission events than normal enriched U02 fuel at similar burnup and should show a lower amount of densification. The klestinghouse

11.

experience with densification of HO fuel operated in a corrmercial power X

reactor is compared with predicted values in Figure 1 which is taken from MCAP-8349-P. The Figvre also contains data for the perfonnance of a gg2 fuel with similar fabrication characteristics for direct comparison.

Figure 1 contains data for both mechanically mixed and "mastermixed fuel. Hastermixed fuel is prepared by co-precipitating UO2 and:Pu02.,

then mixing the Co-precipitated material with U02.

The U02 model is a best estimate model and hence there should be as many 9

,points with densification under predicted as overpredicted. This is obviously not the case for either the U02 or HO fuel with the densification X

being generally overpredicted. Significantly, the prediction of'he model was noticeably more conservative for. the HO fuel than for the U02 with X

similar processing history. As shown in Figure 1, the prediction for master mixed HO was less conservative than for mechanically mixed HO . This is as X X

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expected and further illustrates the lower densification expected from fuel

'ith inhomogenously distributed fission events.

s The performance of an HO . fuel of higher Pu02 content which was operated in X.

the Saxton test reactor was also reported in iiCAP-8349-P . This fuel, which

.operated at higher temperatures than commercial power reactors, was also found to densify less than predicted by the model.

s The predictive capability of the U02 model was further tested by evaluating the performance of the HO fuel in the EPRI Ht0 densification program. These X X fuels were prepared by a wide variation in processing variables and hence represent a range of conditions greater than was present. in data on which the model was based. The performance of the EPRI fuels as reported in EPRL HP-637 are compared with model predictions in Figure 2. The data in Figure 2 indicate the model is essentially conservative or all fuels shown, even those unstable fuels prepared by procedures dissimilar to those for fuel on whose performance the model is based. For the stable fuels, those which densified < 2 volume percent, the model is highly conservative in that it greatly overpredicts the densification. The Ginna HO fuel is X

projected to behave stably; all currently .manufactured >lestinghouse U02

. fuels are also projected to show stable behavior.

12 0

Reg. Guide 1.126 does not indicate a maximum amount of densification allowable but, rather defines the amount of densification that must be assumed based on a thermal resintering test. Oesign must account for the amount of densification assumed from the resinter test. Thermal resinter data is not available for the Ginna HO fuel so an alternate X

method must be chosen to predict the densification performance; the method used was the Westinghouse U02 model. The accuracy of this model can be compared to the predictions from the thermal resinter model in Reg. Guide 1.126.

The basis for establishing the relative abilities of the two models to predict performance was established by use of the EPRI data. The predicted performance using the thermal resintering model is compared with actual performance in Figure 3. Examination of data in Figure 3 indicates the P

resinter model significantly overpredicts the amount of densification, especially in the .region of stable fuels (< 2 volume percent densification) where the overprediction is as much as 2.5 percent; Comparison of Figures 2 and 3 indicates the relative predictive abilities of the two'odels. The comparison shows the two models are very s'imilar, both models being generally-conservative. The degree of conservatism is found to be greater in the region of stable fuels. Similar numbers of data points are'lightly under-predicted by the two models, however, the Westinghouse model'is both more consistent and more conservative in the less than'2 percent densification

'ange which is representative of the Ginna MO fuel.

X The data presented demonstrate that the Westinghouse U02 fuel: densificatign model yields predictions similar to'nd more reliably conservative for YO X

fuel than the resinter model in Reg. Guide 1.126.

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Figure 2. Westinghouse UO> ltodel Prediction of Densification of EPRI HO X

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17 R~es onse The mixed oxide fuel mechanical design is equivalent to the design of Ginna fuel regions 7, 8 and 9. These regions were delivered for insertion into the reactor in 1975, 1976 and 1977, respectively. All Region 7 assemblies have been discharged from the reactor. Peak assembly burnup was 31456 MWD/MTU. Region 8 and 9 fuel assemblies are currently in the reactor. Region 8 assemblies will be discharged during the spring 1980 refueling outage and will not be reinserted. We expect peak assembly burnup for Region 8 in excess of 34,000 MWD/MTU. Region 9 fuel will be reinserted for Cycle 10 with a peak assembly burnup at the beginning of Cycle 10 of nearly 25,000 MWD/MTU.

Performance of these assemblies has, in general, been excellent.

Monitoring of the primary coolant activity has not indicated any fuel failures in these fuel assemblies. A program of visual examination of fuel has been performed for Region 8 fuel not currently in the reactor. This program was initiated because of the results of a fuel examination at another reactor earlier in 1979. That examination indicated that rods from a particular ingot used for zircoloy tubing displayed rod bowing. Since that ingot had also been used for some rods in Region 8 fuel, fuel assemblies containing rods fabricated from that ingot as well as other fuel assemblies were reviewed. It, was determined that rods from that ingot did display some bow but, that there was no bowing discernible in rods not fabricated from that ingot. The bowing was within licensed limits. Further, no other anomalies were seen in the fuel. Peak assembly burnup for the fuel assemblies which were examined was 29067 MWD/MTU. It should be noted that, none of the tubing for the mixed oxide fuel was fabricated from the subject ingot.

Flux traces from the incore detectors are obtained and examined at regular intervals." These also indicate excellent fuel per-formance.

Based on the coolant activity history, on the visual examinations and on the incore detection monitoring, we expect excellent performance from the mixed oxide fuel.

UESTION 10 Table 5.2 of XN-NF-79-103 shows that the control rod worth is less for Cycle 10 than for the previous cycle. Is this effect due to the presence of the mixed oxide fuel? How was this effect included in the accident analyses for Cycle 10 operation.

R~es onse The primary effect resulting in the reduction of the Cycle 10 control rod worths over that of previous cycles is due to the control rods being positioned in assemblies having exposures greater than 16,000 MWD/MT. In previous cycles some control rods

1 I

'I were positioned in'assemblies with exposures of about 5,500 to 6,700 MWD/MT, thus increasing the effective worth. of the control bank. The effect on control rod worth due to the presence of the MOX assemblies is assessed to contribute only minor effects. As reported in XN-NF-79-103, the calculated control rod worths still provide adeguate shutdown margin. The reduced control rod worths are still bounded by the previously used PTS (XN-NF-77-40, Rev.

1) values of 1,890 pcm at BOC and 2,830 pcm at EOC. The corre-sponding values reported in XN-NF-79-103 are 4,807 pcm for BOC and 5,126 pcm for EOC.

UESTION 11 It is in MOX known that the uncertainties associated with power distribution fuel assemblies are in general greater than those in UO2 fuel assemblies. Was this effect considered for the four MOX fuel assemblies to be irradiated in Ginna for Cycle 10?

~Res onse The uncertainties associated with power distributions in MOX fuel assemblies were considered with respect to loading the 4 (four) assemblies in the low power region of the core (the core periphery).

This enhances the margins with respect to the Technical Specification limits and the peak limiting assembly thus allowing for an increased uncertainty.

Discuss the difference in calculating the effect on F , peak heat flux factor, of fuel rod bowing since the calculationS were done for U02 fuel.

~Res onse The Ginna fuel has not displayed the tendency to bow as has been observed in other fuel. This is attributed to the specific fuel design used at Ginna. The fuel assembly utilizes stainless steel guide tubes and has 9 spacer grids. This is as opposed to other designs which have zircaloy guide tubes and only 7 spacer grids.

As described in our letters of August 18, 1976 and February 11, 1977 and as elaborated in our response to Question 9 above, inspections of Ginna fuel demonstrate that the phenomenon noted at other facilities are not observed at Ginna.

The local power increases in MO fuel due to bowed rods is greater than that calculated for U02 full. Based on calculations on the effect of MO2 fuel on power spikes, the qualitative assessment of Westinghouse is that although with a given amount, of bowing, MO2 fuel will have a longer bowing power spike than will UO fuel, this is offset by the fact that. Ginna fuel has less bowing than assumed in the Westinghouse analytical model. Therefore, the statistical combination of power peaking factor uncertainties

("mini-convolution" ), as accepted by the NRC, results in a total power peaking factor uncertainty for Ginna MO fuel consistent with the current Technical Specification limits.