ML18141A605
| ML18141A605 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna, 05000000 |
| Issue date: | 04/09/1984 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML18141A604 | List: |
| References | |
| NUDOCS 8404260444 | |
| Download: ML18141A605 (4) | |
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e UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION RELATED TO EXPENDED FUEL BURNUP AT SURRY AND NORTH ANNA POWER STATIONS Introduction By letter dated November 23, 1983 (Reference 1) Virginia Electric and Power Company (the licensee) requested that the fuel burnup restriction in the Safety Evaluation supporting Amendment Nos. 73, 74, 36 and 16 for Surry and North Anna Power Stations be removed.
These amendments were issued on January 19, 1982.
Although the licensee submitted information (References 2-4) including a general evaluation for burnups to 45,000 MWD/MTU, the staff limited fuel burnup to 37,000 MWD/MTU until further review was done.
The burnup proposed by the licensee is 45,000 MWD/MTU.
Our review is contained herein.
Although the request was for the removal of the restriction, we feel that it is important to leave the restriction but raise the limit to 45,000 MWD/MTU.
Evaluation The licensee's letter dated November 23, 1983, cited a Westinghouse topical report, WCAP-10125 (Reference 5) entitled "Extended Burnup Evaluation of Westinghouse Fuel".
This report describes the models and methodology used in the safety analysis of Westinghouse fuel at extended burnup and discusses the experimental data used to support those models.
Our review of this topical report has not yet been completed.
However, the following observations can be made:
- 1.
WCAP-10125 not only discusses models, methodology and data, but has also applied these models to show that existing limits continue to be met over a burnup range exceeding that requested by the licensee.
- 2.
The models used have been previously reviewed and approved by the NRC without explicit burnup limits.
The analysis simply applies these unchanged models over a burnup range not previously considered, but not including radiological aspects which are discussed below.
- 3.
Westinghouse has examined the application of the existing methodology at extended burnup and has identified no burnup-dependent phenomena which would invalidate the analysis performed.
- 4.
Results of Westinghouse extended burnup Lead Assembly programs at a number of Westinghouse plants (including Surry and North Anna References 6 and 7) support the Westinghouse conclusion (see Item No. 3), but not including radiological aspects which are discussed be l ow*
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8404260444 840409 PDR ADOCK 05000280 p
- 5.
- 6. The licensee has reviewed the Westinghouse report (WCAP-10125) and has determined that the results are applicable *
. Our generic review of WCAP-10125 is expected to be completed in June 1984, whereas it is the licensee 1s intent to gradually increase the average discharge exposures (i.e., over a 3-4 cycle period). This will allow adverse staff conclusions (if any) on WCAP-10125 to be applied prior to achieving the target discharge burnups.
We have performed a preliminary review of the radiological consequences of accidents as discussed in the Westinghouse report and have concluded that the analyses presented are 11best estimate 11 models and do not contain suitable conservatism as appropriate for consideration of accidents. Therefore, we have performed an evaluation of accidents which does provide the required measure of conservatism. There are five factors that must be considered, since they affect the radiological consequences of accidents for extended burnup.
They are (1) changes in the fuel failure rate, (2) changes in the total inventory and mix of radioisotopes in the fuel, (3) changes in the fraction of the isotopes accumulated in the fuel-clad gap, (4) iodine spiking behavior, and (5) the effect of the fuel rod gas pressure on decontamination factors (DF) assumed for fuel handling accidents.
The first of these factors, the failure rate of the fuel rods during accidents, is reviewed as part of the fuel behavior and thermal-hydraulic analyses and is addressed as part of the review above.
The analyses provided in the amendments issued January 19, 1982 are still valid for this requested change for the second and the last two factors (the mix or radioisotopes, the iodine spiking behavior, and the DF).
The final factor, the accumulation of volatile radioisotopes in the gap, is discussed below.
In considering those accidents where the radioisotope content of the gap is important, we have evaluated whether or not the traditional assumption concerning the gap fraction of volatile radionuclides (10%, except Kr-85 which is 30%) remains valid.
Our evaluation utilizes two 11best estimate 11 assumptions, namely, the volatile fission product release model in the ANS 5.4 standard (Reference 8), and the fuel managment scheme for past and present reloads at the four plants.
The scheme is based on locating higher burnup modules in non-limiting locations and filling limiting locations with either fresh fuel or fuel at the beginning of the second cycle.
The increase to 45,000 in the batch average at discharge burnup represents an increase in the burnup of the limiting locations at the limiting time in cycle, to about 16,000 MWD/MTU.
This value has been assumed in this analysis.
Suitable conservatism has been maintained by assuming the plant-specific maximum linear heat generation rate (LHGR) at the plants: 11.4 KW/ft at North Anna and 13.5 KW/ft at Surry.
These values are based on the respective Technical Specification maxima for peaking factors.
For North Anna, the results of the analysis show that the traditional assumptions concerning *gap content are not exceeded for any radionuclide.
- - For Surry, however, the analysis shows that more than 10% of th I-131 is predicted to be contained in the gap, but significantly less than 10% is predicted for all other radioiodines and for the noble gases (except Kr-85, which remains less than 30%).
This is important only for the fuel handling accident, because for fuel failures taking place during accidents occurring during core operations the assumption of 10% of each radioiodine nuclide in the gap will still provide a limiting calculation.
I-131 contributes only about half of the dose to the thyroid in these cases.
For the fuel handling accident, which is postulated to occur 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown, the thyroid dose is dominated by the I-131 release because the other radioiodines of importance have decayed significantly from their levels during operation.
The ANS 5.4 model, combined with Surry's peak LHGR, burnup of the limiting module, and axial power shape, predicts about 13% of the rod 1s iodine inventory in the gap.
We evaluated the fuel handling accident in Reference 9 and determined that the thyroid dose at the Exclusion Area Boundary (EAB) was 34 rem.
Correcting this value to account for the increased release of I-131 relative to the traditional assumption yields a thyroid dose at the EAB of 44 rem.
This value is less than the guideline value of 75 rem in the Standard Review Plan and shows that the design of the plant is adequate to mitigate the consequences of this accident.
Conclusions We have concluded that operation of Surry Units 1 and 2 and North Anna Units 1 and 2 to 45,000 MWD/MTU batch average at discharge using 4.1 w/o enriched fuel is acceptable.
Implicit in this evaluation are the assumptions that there will be no increase in the peak linear heat generation rate and that the fuel management scheme continues to provide the limiting location in
- fuel at the beginning of the second cycle of irradiation.
Date:
April 9, 1984 Princi1 a1 Contributors:
J. Vog ewede, CPB J. Mitchell, AEB D. Neighbors, ORB#l REFERENCES
- 1.
W. L. Stewart (VEPCO) letter (Serial No. 678) to H. R. Denton (NRC) on 11 Extended Burnup Operation to 45,000 MWD/MTU Batch Average Burnups 11 dated November 23, 1983.
- 2.
B. R. Sylvia (VEPCO) letter (Serial 195) to H. R. Denton (NRC) on "Proposed Technical Specification Change 11 dated March 26, 1981.
- 3.
H. R. Leasburg (VEPCO) letter (Serial No. 432) to H. R. Denton (NRC) on 11Supplementary Information on Surry Power Station Units No. 1 and 2, 11 dated July 24, 1981.
- 4.
H. R. Leasburg (VEPCO) letter (Serial No. 495) to H. R. Denton (NRC) on 11Application for Withholding Proprietary Information from Public Disclosure," dated August 18, 1981.
- 5.
P. J. Kersting, Ed., "Extended Burnup Evaluation of Westinghouse Fuel,
11 Westinghouse Electric Corporation Report WCAP-10125 (Proprietary), July 1982.
- 6.
H. D. Moss, Comiler, "Extended Fuel Burnup Demonstration Program, Semiannual Technical Progress Report for the Period from Janury 1983 to June 1983, 11 Westinghouse Electric Corporation Report WCAP-10389 (DOE/ET 34104-12), November 1983.
Prepared for the U. S. Department of Energy by Westinghouse Electric and Virginia Electric and Power Company.
- 7.
S. K. Kapil and R. D. Ankney, "Extended Fuel Burnup Generic Technical Studies," Westinghouse Electric Corporation Report WCAP-10414 (DOE/ET 34014-9), December 1983.
Prepared for the U. S. Department of Energy by Westinghouse Electric and Virginia Electric and Power Company.
- 8.
11American National Standard Method for Calculating the Release of Fission Products from Oxide Fuel", ANSI/ANS~5.4, 1982.
- 9.
Safety Evaluation by the Division of Reactor Licensing, U. S. Atomic Energy Commission in the Matter of Virginia Electric and Power Company Surry Power Station Units 1 and 2, February, 1972.