ML20011A438
| ML20011A438 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 05/18/1976 |
| From: | Madgett J DAIRYLAND POWER COOPERATIVE |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20011A420 | List: |
| References | |
| LAC-3929, NUDOCS 8110130352 | |
| Download: ML20011A438 (15) | |
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N.SIIt)LtNIb FOWEIt CONFEIt.tTIVE
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34601 May 18, 1976 In reply, please refer to LAC-3929 DOCRET NO. 50-409 Director of Nuclear Reactor Regulation Division of Reactor Licensing U.
S. Nuclear Regulatory Commission Washington, D. C.
-20555
SUBJECT:
DAIRYIAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACRWR)
PROVISIONAL OPERATING LICENSE NO. DPR-45 APPLICATION FOR A?iENDMENT TO I.ICENSE
Reference:
(1) 10CFR Part 50, Section 50.90
Dear Sir:
In accordance with,the provisions of Reference 1, an application to amend Provisional Operating License No. DPR-45 for the La Crosse Boiling Water Reactor is hereby filed with three (3) signed original applications together with thirty-seven (37) copies.
This application for amendment to the subject license involves a proposed change in the facility as described in the Safety Analysis-Report.
The proposed change will be the intended use of nuclear fuel which has been designed and fabricated by the Exxon Nuclear Company, Inc.
This fuel is identified as Type III and its first use is proposed for Cycle 5 which is estimated to commence in May 1977.
The Atomic Energy Division of Allis-Chalmers Mfg. Co. designed and fabricated all nuclear fuel which has been used in the LACBWR.
This fuel '.'s been adequately described in the Safety Analysis Report, ACNP-655<.4, and amendments.
All nucicar fuel currently in the reactor is of Allis-Chalmers design and manufacture and is identified as Type II fuel.
At the end of Cycle 4, twenty-four of the highest ex-posed Type II fuel assemblics will be discharged.
The remaining forty-eight (48) Type II assemblies will b2 rearranged with twenty-four (24) new Type III assemblics to form a mixed core for Cycle 5.
A technical description of the Type III nuclear fuel is included with this application.
This document entitled -
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"DESCRI.PTION OF EXXON TYPE III NUCLEAR PUEL "OR BATCli 1 Il: LOAD IN T!!E Li C!">SSI: BOILING WATER REACTOR ( LACDWR)"
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20110130352 010929
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L DR ADOCK 05000
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Director of Nuclear Reactor Regulation LAC-3929 Division of Reactor Licensing May 18, 1976 L,
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compares design and predicted performance of the Exxon fuel with the Allis-Chalmers fuel presently licensed for use in the LACBWR.
The mechani cal, thermal, hydraulic and nuclear properties of the Type III fuel are discussed.
A preliminary description of the core configur-ation for Cycle 5 is contained in the report together with an eval-uation of several postulated accidenta and transients for the core configuration which combines both Type II and Type III fuels.
The proposed change in the facility has been reviewed and approved by the LACBWR Safety Review Committee.
Proposed changes to Technical Specifications which will become necessary as a result of the initial use of the Type III fuel together with Type II fuel will be submitted shortly.
If there are any questions concerning this application, please con-tact us.
Very truly yours, DAIRYLANC ?OWER COOPERATIVE ee p%
Joh P. Madgett, Gen 1 Manager JPM:REE:af ec:
J.
Keppler, Reg. Dir., NRC-DI.O III STATE OF WISCONSIN )) ss.
COUNTY OF LA CROSSE) d Personally came before me this h '
, day of May, 1976, the above named John J. Madgett, to me known to be the person who executed the for going instrument and acknowledged the same.
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--,,Nlx/!.
J Notary PuLlic, La Crosse County, Wisconsin
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My Commission 1:xpires
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T DESCRIPTION OF EXXON TYPE III NUCLEAR FUEL FOR BATCl! 1 RELOAD IN THE LA CROSSE BOILI!!G WATER REACTOR (LAC 1'WR) l Dair'/ land Power Cooperative 2615 East Avenue South La Crosse, Wisconsin 54601 May 17, 1976
V.-
ACCIDENT AND TRANSIENT ANA'L5 SIS This section provides an evaluation of the effect that
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Type III fuel has on all limiting types of accidents and trans-ients considered for the original fuel.
VA.
LOSS-OF-COOLANT ACCIDENT ANALYSIS Loss-of-Coolant Accident Analyses have been performed for the LACBWR Exxon Nuclear reload fucl.
The analysis was performed using as input the syste.? analysis results as previously provided and apprcved.IIII4)
The Exxon Nuclear Company Evaluation Model for Boiling Water Reactor Loss-of-Coolant Accidents ( }
as applicable to LACBWR was used
".n these analyses.
The fuel-to-cladding gaps and gap coefficients were evaluated as a function of assembly ex-posure using GAPEX which was discussed in Scction III.B.3.1.
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l The evaluation model used in the analysis
+;uires explicit data as applicable to stainless steel clad fuel.
Modifications were made to the multirod heatup code liUXY I3I to calculate cladding temperatures during the LOCA.
These modifications were:
adding new rod type options in !!UXY, along with supplying the correct material properties for the cladding.
The properties added for stair.less steel included heat capacity versus temperature, thernal conductivity versus temperature, ar.d a representation of the netal-waier---
water reaction.
The above iter:s are explained in detail in ref erence 3.
i In performing the rod heatup calculations, the following f
general assumptions were made.
The blowdown information for the cladding temp.rature 4
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limiting case as given in Reference 1 shall apply.
This corresponds to the intermediate break size of 0.20 to 0.75 ft2, No rod ballooning or failure is assumed based on the data and information presented in References 1 and 4.
The effects of fuel densification shall be incorporated in the analysis.
All red heatup calculations shall be peric:mcd at 10'%
of full power.
100i cf the power generated was assur.ed to occur in the fuel.
The car wetting is calculated'for each transient internal to HUXY using the =cthod specified in Appendix K. (5)
The blowdo'in and spray heat transf er coef ficient ir. formation used in the rod heatup calculations appear in Figure VA.1 through VA.3 and Table VA.l.
The data for the above figures and table were obtained from the information contained in Reference 1.
The rod types listed in Figure VA.2 are defined sucn that the Type 1 rods are the outermost rods, Type 2 rods are one row from the outside, etc.
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t TABLE VA._1 EVE ;i IIMiS F0P. ROD H:ATil? CALCt1 Ail 0 b Eyent_
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la using the above model and blowdo.;n data, the peaking factors t;are adjusted such that the hattest rod ww oparating at the design pea!.ing These analyses are thus perfor:ced for the storst limit throughout life.
expected operating conditions at each exposure.
The resuits of the analysis are shown in Figa e VA,i and VA.S.
Figure VA.4 shows tne cladding te. perature of selected rods as m
well as the shroud te?,nerature through the transient i.t beginning-of-life.
The corner peak rod and the peak clad te.gerature along with the shroad behavior through th2 transient are sho.;n in Figure VA.4.
Figure VA.5 and Table VA.2 show the peak cladding temperature as e>.perienced ducir.g a !.GCA es a functica of asse:: Sly exposure, using tha assen.bly p<r.iar distributions fra:. Sec tion IIIC. The resul ts sh.. i.ha t for the worst case at each exposur e that the peak clacding temperattare rer,a ins below 2T;3' F throu ;:wa t as,c::.bly li fe.
In addition to calculatinj the peak cladding temperature it was deaonstr. ted th:.t less then 1% of the total cladt '.q thickness in f
the entire ccre rear.t; durinj tie transient.
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fia axial dapendence cr cladd:ng rcactica was assucad.
li v,.e se r, it was assm. J that tne clad reactcd at every cxial t osition to he c(tent that it reacted at the plane or interest.
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recction is a f'.:.nion of asset:bly radial peak i r,"; f ac ;r,r l
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core.
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REFERENCES FOR SECTION V 1.
" Technical Evaluation - Adequacy of La Crosse Boiling Water Reactor Emergency Core Cooling System", Gulf United Nuclear Fuals Company, SS-942.
2.
" Exxon Nuclear Evaluation Model for BWR Loss-of-Coolant Accident", Exxon Nuclear Company, XN-235, W. S. Nechodom and L.
II. Steves, October, 1974.
3.
"IIUXY:
A Generalized Multirod Heatup Code", Exxon Nuclear Company, XN-73-34, Revision 2, L. 9.
- Steves, S.
F. Gaines and J.
D.
Kat:n, October, 1974.
4.
Response to Questions by AEC/DL with regard to Gulf United Report SS-942, SS-1075, Revision 1, November 15, 1973 as forwarded by LAC-2106, January 17, 1974.
5.
" Loss-of-Coolant Accident and Emergency Core Cooling Models for General Electric Boiling Water Reactors", General Electric Company Report NEDO-10329, April, 1971.
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