ML20235K673
| ML20235K673 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 09/29/1987 |
| From: | Frank Akstulewicz Office of Nuclear Reactor Regulation |
| To: | Mroczka E CONNECTICUT YANKEE ATOMIC POWER CO. |
| References | |
| TASK-2.K.3.30, TASK-TM TAC-45830, NUDOCS 8710050192 | |
| Download: ML20235K673 (5) | |
Text
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September 29, 1987 l
1 Docket No.: 50-213 DISTRIBUTION Docket File.
EJordan NRC/ Local'PDR JPartlow Mr. Edward J. Mroczka, Senior Vice President ISAPD File TBarnhart (4)
Nuclear Engineering and Operations DCrutchfield WJones Connecticut Yankee Atomic Power Company FSchroeder ACRS(10)
Post Office Box 270 MShuttleworth Tech Br-i 3
Hartford, Connecticut 06141-0270 FAkstulewicz DHagan 0GC-Beth WJones
Dear Mr. Mroczka:
EButcher GPA/PA ARM /LFMB
SUBJECT:
SAFETY EVALUATION FOR NORTHEAST UTILITIES FUEL R0D ANALYSES PROGRAM (NUFRAP) - (TAC N0. 45830)
Re:
Haddam Neck Plant By letter dated July 24, 1983, Connecticut Yankee Atomic Power Company (CYAPC0) 1 submitted a report entitled "NUFRAP:
Northeast Utilities Fuel Rod Analysis Program" for staff review and approval. By letter dated November 7, 1986, CYAPC0 further stated that the application of the NUFRAP code will be used only to provide fuel rod data for small-break loss-of-coolant accident (LOCA) initial conditions.
If CYAPC0 intends to use the NUFRAP code for other licensing applications, CYAPC0 would provide additional documentation for NRC review and approval.
The staff has completed its review and has concluded that the NUFRAP code is acceptable for use in calculating initial conditions of fuel temperature and fission gas release for small-break LOCA analyses. However, this approval does' not extend the use of NUFRAP to other fuel performance analyses, e.g., transient j
conditions, since the code has not been reviewed and approved for these other applications. A copy of our related Safety Evaluation is enclosed.
This completes all staff review on this issue.
1 Sincerely, I
original signed by Francis M. Akstulewicz, Jr Project Manager Integrated Safety Assessment i
Project Directorate Division of Reactor Projects III/IV/V and Special Projects
Enclosures:
As stated cc: see next page i
V' FAkstulewcz:/
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g /g/87 g)/87 0FFICIAL RECORD COPY 8710050192 870929 DR ADOCK 050 3
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Mr. Edward J. Mroczka Haddam Neck Plant Connecticut Yankee Atomic Power Company cc: Gerald Garfield, Esquire Kevin McCarthy, Director Day, Berry & Howard Radiation Control Unit Counselors at t.aw Department of Environmental City Place Protection Hartford, Connecticut 06103-3499 State Office Building Hartford, Connecticut 06106 Superintendent Richard M. Kacich, Manager Haddam Neck Plant Generation Facilities 1.icensing RFD #1 Northeast Utilities Service Company Post Office Box 127E Post Office Box 270 East Hampton, Connecticut 06424 Hartford, Connecticut 06141-0270' Wayne D. Romberg Donald O. Nordquist, Director Vice President, Nuclear Operations Quality Services Department Northeast Utilities Service Company Northeast Utilities Service Department Post Office Box 270 Post Office Box 270 Hartford, Connecticut 06141-0270 Hartford, Connecticut 06141-0270 i
Board of Selectmen Town' Hall Haddam, Connecticut 06103 Bradford S. Chase, Under Secretary Energy Division Office of Policy and Management 80 Washington Street Hartford, Connecticut 06106 Resident Inspector Haddam Neck Nuclear Power Station c/o U.S. NRC P. O. Box 116 4
East Haddam Post Office I
East Haddam, Connecticut 06423 l
Regional Administrator, Region I j
U.S. Nuclear Regulatory Commission i
631 Park Avenue King of Prussia, Pennsylvania 19406
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gv.....f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO IMPLEMENTAT]ON OF TM1-ACTION ITEM ll.K.3.30
" REVISED SMALL BREAK LOCA ANALYSIS" CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT, DOCKET N0. 50-213
1.0 INTRODUCTION
TMI Action item II.K.3.30 describes a need to revise the small-break loss-of-coolant accident (LOCA) analysis for compliance with Appendix K to 10 CFR Part 50. The revised analysis should be documented and submitted for NRC-l approval. Connecticut Yankee Atomic Power Company responded to the NRC request by submitting a topical report entitled "NUFRAP: Northeast Utilities Fuel Rod Analysis Program". (NUSCO 135-P) dated July 24, 1983 for NRC review.
This topical report documented a steady-t. tate fuel performance computer code NUFRAP for calculating fuel rod mechanical responses including small break LOCA initial conditions for Haddam Neck. NUFRAP was derived from the FRAPCON-2 fuel performance code developed by Pacific Northwest Laboratory and i
Idaho National Engineering Laboratory for NRC. However, because of the un1gue feature of stainless-steel fuel rod cladding, the licensee incorporated j
stainless-steel properties into FRAPCON-2 to describe the in-reactor cladding behavior in Haddam Neck. The licensee compared the analytical results with the available data in the mechanical, fission gas release, and densification I
models.
1 In a letter dated November 7,1986, from J. F. Opeka to NRC, the licensee stated that the NUFRAP code will only be used to provide fuel rod data for str.all break LOCA initial conditions, and for other licensing applications, additional documentation will be provided for NRC review. Our review and i
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2 evaluation are thus limited to the applicability to small break LOCA i
analyses.
2.0 MODEL DESCRIPTION AND EVALUATION As mentioned earlier, NUFRAP is derived from FRAPCON-2, which is an NRC-sponsored computer code for stady-state fuel performance analysis. The licensee retains most of the thermo-nechanicat models except the cladding model which is converted from Zircaloy properties to stainless steel properties.
The licensee provided a list of input options f or mechanical, fission gas release, and fuel densifications models. These options are standard for the FRAPCON-2 code users. Hence we consider that these selections are acceptable.
The stainless-steel model consists of thermal conductivity, elastic modulus, thermal expansion, shear modulus, and creep submodels. These submodels were developed by Hanford Engineering Development Laboratory and are typical for fast breeder reactor cladding, therefore, we consider that the stainless steel cladding model is acceptable for use in NUFRAP.
3.0 COMPARISON OF PREDICTION WITH DATA The licensee compared the analytical results with post-irradiation examination data of temperature and fission gas release. There are ruch fewer post-irradiation examination data of stainless-steel cladding fuel rods than the data of Zircaloy cladding fuel rods. Hence it is difficult to evaluate
- However, the adequacy of NUFRAP based on the limited number of data points.
since the licensee indicated that the current goal of NUFRAP is for small break LOCA initial conditions and the only two pertinent quantities for LOCA analys1s are stored energy (fuel temperature) and rod pressure (fission gas I
release), it is appropriate to address conservatism rather than best estimation for evaluating the adequacy of NUFRAP.
i
-c 3
_ (a) Fuel Temperatur_e There were -no direct fuel temperature measuremen s available for t
stainless-steel clad fuel rods.
The licensee deduced the fuel temperature from grain structure of irradiated fuel pellets. This type of deduced fuel temperature usually is associated with large uncertainties. The licensee The presented only two such data points for comparing with the predictions.
results showed that the. predicted temperatures are much higher than the deduc'ed temperatures. Although' there are only two temperature data points available, the NUFRAP code shows adequate conservatism in fuel temperature prediction.- We therefore conclude that the fuel temperature predictions using NUFRAP is adequate for small break LOCA initial conditions.
-(b). Fission Gas Release 1
High fission gas release results in high internal rod pressure for LOCA initial conditions. The rod pressure affects the cladding swelling and rupture during LOCA condition. The-licensee presented eight data points for comparing with the NUFRAP predictions. The fission-gas release model in NUFRAP is the ANS 5.4 model, which is accepted by NRC.
The results showed that overall the predictions were higher than the measurements.
Hence we-conclude that the fission gas release prediction using the ANS 5.4 model in NUFRAP is conservative and thus acceptable for small break LOCA initial j
conditions.
4.0 CONCLUSION
Based on the conservatism shown in comparing the analytical results with experimental measurements, we conclude that the NUFRAP code is acceptable for calculating initial conditions of fuel temperature and fission gas release for use in small break LOCA analyses.
However, this approval does not automatically utend the use of NUFRAP to other fuel perfonnance analyses, I
e.g., transient conditions since the code has not been reviewed and approved for these other applications.
5.0 ACKNOWLEDGEMENT Principal Contributor: Shih-Liang Wu, SRXB, NRR
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