ML19327B980
| ML19327B980 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/01/1989 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | Swenson W NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| CON-NRC-89-135 TAC-68862, TAC-68863, VPNPD-89-572, NUDOCS 8911140331 | |
| Download: ML19327B980 (7) | |
Text
{{#Wiki_filter:t 4-Wisconson Electnc u l MWER COMPANY 231 w Mch9m.Po tve 2cn hoAsa # !3?oi u)22523e 1 + VPNPD-89-572 +NRC-89-135 November 1, 1989
- i Document Control Desk r
i U. S. NUCLEAR REGULATORY COMMISSION Mail Station PI-137 i Washington, D.C. 20555 A?.tention: Mr. Warren Swenson, Project Manager Project Directorate III-3 i GLntlemen DOCKET NOS. 50-266 AND 50-301 TECHNICAL SPECIFICATION CHANGE REQUEST NO. 124 NUCLEAR FUEL STORAGE ENRICHMENT (TAC NOS. 68662 AND 68863) U POINT BEACH NUCLEAR PLANT Your letter dated October 13, 1989, requested additional information regarding our license amendment application dated July 6, 1988, 5 concerning. fuel storage enrichments at our Point Beach Nuclear Plant, d Unit Nos. 1 and 2. The attachment to this letter contains our response to each of your five requests. You should be aware that we have ordered fuel for the Point Beach Unit 1 spring 1990 refueling with nominal enrichments of up to 4.2 weight percent uranium-235. We request that our amendment application be approved by February 1,.1990 so that we may receive and store the new reload fuel as it is reccived from the vendor. Please contact us if you have any additional' questions concerning the applicution or the information provided herewith. Very truly,yours, ? b ~ (h56/ g" C. H. Fa c' ; Vice President Nuclear Power Copies to NRC Regional Administrator, Region III NRC Resident Inspector l '1 R. S. Cullen, PSCW 8911140331 891101 [ccl ADOCK05000gg6 DR dwAsWim ofliinnenDwxy(h mthn 7
ATTACHMENT 1. Describe the correctivo' actions to be taken if degraded Boraflex specimens or absorber is found in the spent fuel pool. Certain types of corrective actions, e.g. use of poison rods or checker-i board loading patterns, could require revised criticality i analysis. [ f
Response
Wisconsin Electric's July 6, 1988 amendment request discusses the results of the criticality analyses of the Point Beach new and [ spent fuel racks for the storage of Westinghouse OPA fuel. These l analyses were performed by Pickard, Lowe, and Garrick, Inc. .(PLG), In the submittal it was noted that the PLG analyses assumed tnere was unborated water in the spent fuel pool. In addition to this, a supplemental case was examined with unborated i water in the pool and no Boron-10 in the Boraflex neutron poison material. This analysis determined that if the fuel storage racks were divided into two regions, with Region 1 using a i checkerboard pattern for the storage of new fuel or fuel with low l burnup and Region 2 utilizing every cell to store spent fuel, a K-infinity of 0.881 would be achieved when new fuel with 4.75 w/o enrichment is stored in Region 1. This information was obtained from the attached eigure 2 from the PLG supplemental report. l This analysis also determined the minimum burnup required for spent fuel stored in Region 2 to ensure that K-infinity remains i below 0.95. The maximum combined bias and uncertainty were determined, an allowance was.made for interpolation error, and the curve shown in the attached. Figure 25 from the PLG report was produced. This shows the minimum burnup for 4.75 w/o OFA fuel to l be stored in Region 2.is 38,400 MWD /MT. Average discharge burnups of 45,000 MWD /wT are planned ~for the higher enriched fuel assemblies. Based on this information,.if degraded Boraflex were found in the spent fuel pool, the storage of spent fuel would not ) be affected, but new fuel or fuel with a burt.;p less than 38,400 I MWD /MT would have to be stored in a designated area in the spent fuel pool using a checkerboard pattern.- We would note also that { j our letter to the NRC' dated April 13, 1989 has proposed a revised j L boraflex surveillance program for the Point Beach spent fuel pool. This program will detect indications of degraded Boraflex-performance and permit us to take the necessary corrective action. 2. Do the Boraflex sheets extend to cover the full length of the active fuel zone, even after maximum shrinkage due to I trradiation? If not, what effect does the resulting gaps have on the criticality analysis?
Response
E The nominal length of the Boraflex sheets is 147 inches which extends beyond the 144 inch length of active fuel in the fuel l assemblies. Our Boraflex surveillance to date, which included a I
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hh1 1 ? ac, i the destructive examination of two panels, has detected no length shrinkage of the Boraflex. This included a panel with Icw gamma exposure and one that had freshly discharged fuel.placed nyyt to i it every six months for six years for a total dose of 1x10 rads p gamma. This is equivalent to the accumulated dose in 20 years l for the average spent fuel rack location. l i. The Electric Power Research Institute (EPRI) completed a study of Boraflex in 1988. They concluded that shrinkage increases as a functionofgammaradiat{gndose,buttheeffectsaturates L between9yf0 and 1.5x10 rads-gamma. The Point Beach panel a with 1x10 rads gamma showed no noticeable length shrinkage. We would note, however, that the amount of shrinkage can vary from L one Boraflex batch to another. The EPRI study concluded that length shrinkage could be as high as 3-4%. This means that in-i j-the extreme case Point Beach panels could end up as short as 141 L inches. As indicated in the response to question #1 above, such l [ shrinkage would have no significant impact on the' criticality analyses conducted for the spent fuel storage racks. l l 3. .The radiological consequences of a spent fuel handling accident need to be analyzed at the higher fuel burnup levels planned for the new higher enriched fuel. Provide the results of this
- analysis, j
Response
fuel The change in the radiological consequences of a spent handling accident due to the increase in fuel burnup levels has been addressed in the approved Westinghouse Electric Corporation document WCAP-10125-P-A, " Extended Burnup Evaluation of p Westinghouse Fuel", December 1985. This WCAP was approved by the NRC for referencing in' license applications in a letter from Mr. .t Cecil O. Thomas.to Mr. E. P Rabo dated October 11, 1985. The l evaluation concludes, "... the results presented in safety analysis reports for offsite dose consequences would not change significantly as a result of the extended burnup." g I i An analysis was performed by Westinghouse at burnup values of j; 33000 MWD /MTU and at 48000 MWD /HTU. The results of these { c analyses show a change in core inventory of between +5% and -13% Lc for short lived nuclides while Kr-85, the only long half-life { radionuclide of significance in calculating release consequences, i exhibits a nearly linear increase in core inventory. The 33,000 MWD /MTU burnup was the level utilized in the original design basis for the PBNP core. With the current optimized fuel i L< assemblies and the proposed increase in enrichment of the fuel, the new target burnup level will be 48,000 MWD /MTU. l ,4 e In order to model the release of fission products due to a fuel handling accident, the amount of fission products released from the fuel pellet matrix into the gap region must be anar d. Westinghouse has performed this analysis with the fo' v ing i L l results: (1) for nuclides with half-1dfes less than one year the peak gap release fraction occurs at an average burnup of 31000
ge m V MWD /MTU and declines with. extended burnup, and (2) for nuclides i with half-13fes greater than one year the gap release fraction continues to increase with increasing burnup. The increase in the gap activity of long-lived Kr-85 and subsequent potential i off-site release has only a small impact on the radiological consequences.due to the small whole body dose conversion factor l for Kr-85. L Combining the changes in the reactor core inventory with the l changes-in the release fraction for burnup values of 33000 MWD /MTU and 48000 MWD /MTU, the Westinghouse analysis summarizes "... the effect of extended burnup on the radiological impact of a fuel handling accident is to increase the thyroid dose by 4 i y percent...[while) the whole body dose is not affected". In the Point Beach Nuclear Plant Final Safety Analysis Report, i f Section 14.2.1, " Fuel Handling Accidents", two calculations were l performed with relation to off-site consequences for the thyroid dose, each derived from an uncorrected reference source term i based on a set of standard assumptions utilized by the Atomic Energy Commission at the time of Point Beach licensing (FSAR, page 14.2.1-8). (1) A calculation of expected dose is based on the highest l powered fuel assembly in the region to be discharged during i normal refueling. This calculation utilizes Westinghouse t [ design characteristics for the fission product inventory in i the gap. region, calculated peak-to-average power levels for 6 .this assembly, and results of studies to determine the i L' expected spent fuel pool water decontamination factor for [ halogens released. These correction values are then applied to the reference source term to obtain the expected off-site L doses (FSAR, page 14.2.1-10).. l (2) A design basisJdose is calculated using a similar approach [ L but with more conservative correction factors to provide for l margin over the expected value. (FSAR, page 14.2.1-11 and 14.2.1-12) The maximum halogen increase of'4% as given in WCAP-10125-P-A can j c h be directly applied to the thyroid doses currently given in the i PBNP FSAR to obtain new bounding doses for this accident. A -l [E comparison of the original and revised thyroid doses is as l follows: [ Calculation PBNP FSAR Anticipated Higher L Type Thyroid Dose Burnup Thyroid Dose l Expected dose 0.8 rem 0.83 rem j [ Design Basis dose 17.5 rem 18.2 rem The whole body dose at the Fite boundary due to a fuel handling accident release is identified in the PBNP FSAR (page 14.2.1-12) as 10 mrem. Based on the Westinghouse analysis this value will not significantly change due to the storage of extended burnup ^ fuel assemblies. i I _ i
@f l-g ~ i [, --..v r y-E i f The expected and design basis values for thyroid dose and whole i body dose resulting from a fuel handling accident'with higher b burnup fuel assemblies are well within the 10 CFR Part 100 g guidelines of 300 rem thyroid dose and 25 rem whole body dose. Including the slight change in the thyroid dose identified by the r Westinghouse analysis, the total thyroid dose remains small compared to the 10 CFR Part 100 guideline i 4. It is the staff's understanding that your request to utilize an p' optional axial-zoned core-loading scheme refers to the use of a . single enrichment throughout the active portion of the fuel t assembly with axial blankets of natural (non-enriched) fuel above and below the active portion of the fuel assembly. Please e confirm our understanding or provide additional justification. i f
Response
I Your understanding of the axial-zoned core-loading scheme is f [ correct. This design, also known as " axial blankets", presently t L incorporates pellets of natural, non-enriched, uranium in the top and bottom six inches of the fuel bearing regions of all fuel v rods in the fuel assembly. 1 5. The requirements of 10 CFR 51.52(b) for a detailed analysis of the environmental effects of transportation of fuel and wastes to l and from the reactor must be satisfied. WEPCO should either L adopt the NRC staff statement developed for Shearon Harris (53'FR I 30355 as corrected by 53 FR 32322) if applicable or WEPCO should t-provide its own statement under 10 CFR 51.52(b). I ' Response: The referenced Federal Register notice presented an NRC staff assessment entitled,J "NRC Assessment of the En'rironmental Ef fects of Transportation Resulting'from Extended Fuel Enrichment and Irradiation." The staff concluded in.this assessment that the environmental impacts summarized in Table S-4 of.10 CFR 51.52 for the burnup level of:33 GWD/MT.are conservative and bound the corresponding impacts for_burnup level up to 60 GWD/MT and i uranium-235 enrichments up to five percent by weight. We have i ~ p reviewed this assessment and conclude that the parameters of the l assessment bound the proposed burnups and fuel enrichments j presented in our application. Accordingly, we would adopt the NRC staff statement and evaluation of the environmental effects [l - of transportation of fuel and wastes developed for the Shearon Harris Nuclear Power Plant at 53 FR 30355 as corrected by 53 FR [ 32322 as fully applicable to the Point Beach Nuclear Plant amendment application dated July 6, 1988. i .}}