ML20072L582
| ML20072L582 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 08/15/1994 |
| From: | Tuckman M DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9409010024 | |
| Download: ML20072L582 (12) | |
Text
ll l
l F
DukaIbwerCompany M S hoans P.O. Box 1006 Senior Vice President Oarlone, NC2820M006 NuclearGeneration
)
(704)382-2200 OMice (104)3824360 Fax DUKEPOWER August 15,1994.
U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Document Control Desk
Subject:
McGuire Nuclear Station Docket Nos. 50-369 and -370 Responses to Request for Additional Information Technical Specification Amendment to Increase Spent Fuel Enrichment Limit
Reference:
Technical Specification Submittal dated June 13,1994 McGuire Spent Fuel Enrichment Limit Upgrade Gentlemen:
Enclosed is Duke Power Company's response to your Request for AdditionalInformation (RAI) concerning the referenced technical specification submittal. The request was recieved by letter dated August 10,1994. We hope you find our responses are sufIicient to satisfy your concerns.
We appreciate your prompt and thorough review of our proposal. Ifyou have additional questions or need additional information, please contact Judy Twiggs at 704-382-8897.
Sincerely, L b-f MMA.
1 M.S. Tuckman Senior Vice President Nuclear Generation jgt/ attachments 94o9010024 940815 PDR ADOCK 05000369 P
PDR l
,,, _ n.
6
=
I
xc:
S.D. Ebeneter, Regional Administrator U.S. Nuclear Regulatory Commission - Region II 101 Marietta Street, NW - Suite 2900
' Atlanta, Georgia 30323 R.A. Martin, Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 14H25, OWFN Washinton, D.C. 20555 i
G.F. Maxwell Senior Resident Inspector McGuire Nuclear Station Dayne Brown. C;ci State of Nctm Carolina -
Division of Radiation Protection P.O. Box 27687 Raleigh, N.C. 27611-7687 1
'I.
l I
l l
ty J
r RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION MCGUIRE FUEL ENRICHMENT INCREASE Q1) Discuss the number of neutron histories accumulated in each KENO Va calculation and why this is considered adequate to assure convergence.
Al) All of the KENO Va calculations used to support this submittal had a nominal 90,000 neutron histories to support the final results. Plots of the avsrage k-effective as a function of neutron generation clearly indicate that the problem has converged to an appropriate solution. Experience has shown that 90,000 histories is more than sufficient to converge most well behaved problems. In addition, with new fuel vault calculations in particular, multiple random number sequence runs were made to confirm that KENO Va had indeed converged to a reasonable answer.
Q2) How do KENO Va calculations with the 123 group GMTH cross sections compare with CASMO-3/ SIMULATE-3 calculations for the same McGuire storage rack configuration?
A2) Calculated reactivities from both CASMO-3 and SIMULATE-3 were used to support this submittal as discussed in Section B.l. Comparisons of the calculated k-infinities between CASMO-3, SIMULATE-3 and KENO Va were performed for both types of the McGuire spent fuel storage racks. The different codes agreed quite well, with the largest difference being 0.00278 Ak. Table A2 below shows these comparisons.
Table A2 Storage Fuel CASMO-3 SIMULATE-3 KENO Va Rack Enrichment k-inf k-inf k-inf Region 1 4.0 0.92053 0.91982 0.92051 Region 2 1.4 0.90229 0.90271 0.90507 Q3) The discussion of fuel mistoading accidents in Section VII.4 of the supporting safety analysis for the license amendment request incorrectly refers to a criticality criterion-of_Keff at or below 0.98. The appropriate criterion should be Kefrat or below 0.95.
Please verify that the 0.95 criterion is met for all misloading accidents with allowed credit for soluble boron.
A3) While the design criterion was incorrectly stated in Section VII.4, all mistoading accidents were verified against the correct design criterion of k g 5 0.95 with e
allowed credit for soluble boron as discussed below. Included in this package is an
update for page 7-5 of Attachment 4 which corrects the criticality criterion for the mistoading accidents.
The effect of a mistoaded assembly in Region 1 was an increase in reactivity of <0.01 Ak. The small reactivity increase for a Region 1 misload is due to the relatively high reactivity of the filler assembly that is replaced by a fresh assembly. The negative reactivity from 2000 ppm soluble boron was worth ~0.17 Ak. Similarly for Region 2, the accidents increased reactivity ~0.06-0.07 Ak and the addition of 2000 ppm soluble boron decreased reactivity ~0.19-0.20 Ak. Hence the allowed credit for soluble boron under these conditions more than compensates for the reactivity increase of a misloaded assembly and the original design criterion of k g.< 0.95 is e
satisfied.
Q4) Since you are proposing to place the boron concentration limit that is maintained in the spent fuel pool in the COLR, the approved analytical methods used to determine this limit must be referenced in the COLR Section of TS 6.9 in order to conform with Generic Letter 88-16. If this has not been done, what are your plans for a TS amendment?
A4) A technical specification amendment package was submitted to you by letter dated May 24,1994 in order to relocate the spent fuel pool boron concentration to the COLR. Supplemental information concerning this request was also provided by letter dated August 4,1994. The McGuire Spent Fuel Enrichment Limit Upgrade technical specification amendment assumes prior approval of the relocation of the boron concentration value to the COLR.
Q5) How does the amount of neutron absorber material in the critical experiments that were used for benchmarking the methodology compare to the amount in the McGuire spent fuel pool storage racks? Explain why any appreciable differences in the amount of absorbing material still makes the benchmarking acceptable.
A5) The absorber material used in the benchmark experiments was slightly borated aluminum sheets ranging from 0.1 to 1.62 weight % of natural boron. The thickness of these sheets was 0.645 cm. The Boraflex absorber material used in the McGuire spent fuel storage racks contained a much higher concentration of boron, but a much thinner sheet. The Boraflex in the Region 1 racks is 34% boron by weight, but only 0.198 cm thick. The Region 2 Boraflex is 30 weight % boron and 0.081 cm thick.
While the amount of absorbing material in the benchmark may not be comparable to that in the storage racks analyzed, it is reactivity that is the primary design criterion, and not the amount of absorber. Thus, a more relative parameter for comparing the different absorbers is the worth, since this is a direct measure of reactivity.
CASMO-3 was used to calculate the worth of the sheets of absorbing material.
t Cases were run in an infinite lattice both with and without the absorbing material.
The calculated worth of the 1,62 wt% borated aluminum sheet in the benchmark experiment was 0.22 Ak. The worth of the Region 1 Boraflex sheet was 0.25 Ak and the Region 2 Boraflex sheet was 0.24 Ak.
3 l
The results of the benchmarking were studied to determine if any trends existed as a function of poison loading. These results are summarized in Figure AS below.
Figure A5 4
0.0050-e u 0.0000
- e-
+---
i 0.25 0.5 0.75 1
1.25 1.5 1.75 i>
W 0.0050 -
-0.0100 Poiwa Leading (wi% th>ron)
Figure A5 shows that no significant trend exists over the full range of poison loading. However, it does suggest that use of an additional bias may be appropriate for the higher loadings such as those found at McGuire. Since the calculated k-effectives overpredict the reactivity for these higher poison loadings, as shown in Figure AS, the decision to ignore this possible trend was conservative.
In summary, although the amount of absorbing material in the benchmark is not representative of the amount of absorbing material in the McGuire spent fuel storage racks, the worths are very similar. Additionally, no consistent trend in Ak exists as a function of poison loading. The benchmark results did suggest that the calculated k-effectives conservatively overpredict reactivity for the poison load:ngs found in the McGuire racks. Therefore, the benchmarking performed to suppoit the CASMO-3/ SIMULATE-3 methodology is acceptable.
l Q6) How much Boraflex shrinkage was assumed in both the width and axial direction?
L What percentage of the total Boraflex dimensions does this represent?
A6) The Boraflex shrinkage in the width direction was assumed to be 0.25" of the 7.5" l
nominal width, or 3.3%. This is consistent with previous coupon examinations and experimental results, and accounts for both shrinkage and edge deterioration.
The axial shrinkage assumption was based on results obtained from in situ blackness testing performed by National Nuclear Corporation in 1991 at the McGuire facility.
i The 45 panels in Region 1 and 36 panels in Region 2 which were examined, had received cumulative gamma exposures of approximately equal to, or greater than, the saturation dose for shrinkage of 10'8 rads. The axial locations of the Boraflex top and bottom ends were determined and compared to both the original as built locations and the location of the fuel stack to determine the length of the fuel stack not covered by the Boraflex. Statistical worst case results were calculated and are shown in Table A6.
Table A6 Exposed Shrinkage at End Fuel Length Region 1:
Bottom 1.34" (l.0%)
4.73" Top 2.07" (1.5%)
3.82" Region 2:
Bottom 0.64" (0.5%)
5.86" Top 1.44" (1.0%)
3.11" In calculating the reactivity penalty for axial Boraflex shrinkage, the worst case exposed fuel length was assumed for both ends. The results of the reactivity calculations showed that there was no change in reactivity for exposed fuel lengths at both ends up to 4.5" for Region 1 and 6.5" for Region 2. Thus, a small reactivity penalty was applied to Region I where the worst case assumed exposed fuellength exceeded the threshold.
Q7) In view of the gaps observed in Boraflex by blackness tests in numerous spent fuel storage racks, why weren't Boraflex gaps considered in the criticality analyses for McGuire?
A7) In order to determine the effect of Boraflex shrinkage and the possible formation of gaps in the Boraflex sheets, testing was performed on the McGuire storage racks in 1991 by National Nuclear Corporation.
As discussed in A6), cells with the saturation dose for shrinkage were examined. These cells should exhibit the largest gaps since gap formation is caused by the radiation induced shrinkage. Results showed only 3 of 79 gaps exceeded 2", but all were less than 3".
This result is consistent with the test results, and the McGuire rack design feature which permits shrinkage to occur at the ends of the Borallex. In studies performed by the storage rack manufacturer, no change in the reactivity of the storage rack is observed until the gaps in the Boraflex sheets exceed I" on all four sides of the storage cell, or 2" on two of the four sides. Therefore, since the gaps found in the Boraflex at McGuire did not exceed this threshold, the formation of gaps in the Boraflex sheets was ignored.
1
Q8) The NRC staff does not agree that the proposed removal of the center-to-center distance between fuel assemblies in the new and spent fuel racks brings these Specifications in line with STS format. Both the current and the improved STS specify these center-to-center distances and they should be retained in the McGuire TS.
A8) We agree that the curren. and improved STS have provisions in the Design Basis Section to specify k g, as well as center-to-center spacing and burnup criteria as e
appropriate for the specific design. We proposed to eliminate information concerning rack spacing since alone, it is not sufficient to ensure that adequate safety margin to criticality is maintained, whereas keff is However, in the interest of maintaining consistency with STS format, we have determined that it is appropriate to maintain 1
information concerning rack spacing in the Specifications. Therefore, enclosed for your review is a modified Specification 5.6.1 containing this information. Please replace the appropriate page(s) of the submittal with this information.
I
9 9
e t
Response to Question 3 l
l
~
All proposed boundary restrictions as discussed above are summarized below:
R_egion Interface Restrictions Region 1 A and 1B Row of region 1B bounding region 1 A must be a row of alternating region 1B fueltypes and region 1B fillerlocations.
Region 2A and 2B No Interf ace Restrictions At least 2 opposite sides of region 20 shall be Region 2A or 2B bounded by either a row of empty water cells or and 2C the fuel pool wall. The remaining side (s) may be either region 2A or region 28 or both.
Region 1 and 2 No interface Restrictions VII.4 Fuel Misloading Accident Analysis The following table summarizes the results of specific analyses performed to verify that sufficient margin is provided by the soluble boron to maintain the pool configuration at or below Keff of 0.95 following various postulated fuel misloading accidents.
Summary of Misloadine Accident Analysis Mistoading Accident Keff Maintained at or Event Description Below 0.95 with 2000 ppm
- Soluble Boron Region 1 " Filler" Location Mistoaded with Yes 4.75 welght% Fuel Region 2 " Filler" Location Misloaded with Yes 4.75 weights Fuel Region 2 Empty Cell Location Misloaded with Yes 4.75 weight % Fuel 7-5
O Attachment II Response to Question 8 l
' Section 5.0 DESIGN FEATURES 5.6 Fuel Storaae CRITICALITY 5.6.1
- a. The spent fuel storage racks are designed and shall be maintained with:
- 1) k er 5 0.95 if fully flooded with unborated water as described in Section e
9.1 of the FSAR; and
- 2) A nominal 10.4" center to center distance between fuel assemblies placed in Region 1; and
- 3) A nominal 9.125" center to center distance between fuel assemblies placed in Region 2.
- b. The new fuel storage racks are designed and shall be maintained with:
- 1) k rr 5 0.95 if fully flooded with unborated water as described in Section e
9.1 of the FSAR; and 2) k,,,5 0.98 if moderated by aqueous foam as described in Section 9.1 of the FSAR; and
- 3) A nominal 21" center to center distance between fuel assemblies placed in the storage racks.
DRAINAGE 5.6.2 The spent fuel storage poolis designed and shall be maintained to prevent inadvertent draining of the pool below elevation 745 ft. 7 in.
CAPACITY 5.6.3 The spent fuel storage poolis designed and shall be maintained with a storage capacity !imited to no more than 1463 fuel assemblies (286 spaces in Region 1 and 1177 spaces in Region 2).
l i
Pronosed Reauirements:
Technical Specification
Reference:
5.6 Fuel Storace Section 5.0 DESIGN FEATURES 5.6 Fuel Storace CRITICALITY 5.6.1
- a. The spent fuel storage racks are designed and shall be maintained with:
- 1) k n s 0.95 if fully flooded with unborated water as described in Section e
9.1 of the FSAR; and
- 2) A nominal 10.4" conter to center distance between fuel assemblies placed in Region 1; and 0
- 3) A nominal 9.125" conter to center distance between fuel assemblies placed in Region 2.
- b. The new fuel storage racks are designed and shall be maintained with:
- 1) k n s 0.95 if fully flooded with unborated water as described in Section e
9.1 of the FSAR; and
- 2) k n s 0.98 if moderated by aqueous foam as described in Section 9.1 of e
the FSAR; and
- 3) A nominal 21" center to center distance between fuel assemblies placed in the storage racks.
DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 745 ft. 7 in.
CAPACITY 5.6.3 The spent fuel storage poolis designed and shall be maintained with a storage capacity limited to no more than 1463 fuel assemblies (2P6 spaces in Region 1 and 1177 spaces in Region 2).
l 8-24