ML20217N013
| ML20217N013 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 08/20/1997 |
| From: | Hughey W ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-96-04, GL-96-4, GNRO:97-00079, GNRO:97-79, NUDOCS 9708260002 | |
| Download: ML20217N013 (4) | |
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August 20,1997 U.S. Nuclear Regulatory Cummission Mall Station P137 Washington, D.C. 20555 Attention:
Document Control Desk
Subject:
Grand Gulf Nuclear Station Docket No. 50-416 License No. NPF 29 Results of Safety Criticality Analysis for GGNS Spent Fuel Storage Racks GNRO:
97/00079 Gentlemen:
Pursuant to your letter of May 14,1997 " Completion of Licensing Action Review For Generic Letter 96-04, 'Boraflex Degradation in Spent Fuel Pool Storage Racks', For Grand Gulf Nuclear Station, Unit 1", we are submitting the required 120 day response (Attachment 1). Your letter requested that we submit the results of the revised Criticality Safety Analysis for the spent fuel pool storage racks.
In your letter, a statement was made wnich is in error. Analysis of Boraflex coupon specimens are not used to determine the size and distribution of gaps.
Specimens are inspected to provide an indication of the general condition of the Doraflex and indicate gross or unusual degradation. Blackness testing is used to determine the sizes and distribution of panel gaps located in the test area.
Additionally, your letter described the use of a 1014 months irradiation time for fuel freshly discharged into the spent fuel pool test area. This is consistent with the Doraflex program developed in response to IN 87-43. This process has been applied for six cycles resulting in the test area obtaining a dose significantly higher than the balance of the pool, in order to avoid excessive dose accumulation in the test area, the response to GL 96-04 proposed an increase in the test interval to 36 months < The irradiation time will be established to ensure the test area leads the maximum panelin the balance of the pool throughout the next test period, f
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if additionalinformation is needed, please contact Riley Ruffin at 601437 2167.
Yours truly, DA
/RR attachment:
- 1. Requestedinformation cc:
Ms. J. L. Dixon-Herrity, GGNS Senior Resident (w/s)
Mr. N. S. Reynolds (w/a)
Mr. L. J. Smith (Wise Calier) (w/a)
Mr. H L. Thomas (w/o)
Mr. Ji W. Yelverton (w/a)
Mr. E. W. Merschoff (w/a)
Regional Administrator
. U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 Mr. J. N. Donohaw, Project Manager (w/2)
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 13H3 Washington, D.C. 20555 8
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GNRO 97/00079 Page 1 of 2
Background
On October 16,1996 Grand Gulf responded to GL 90-04, *Boraflex Degradation in Spent Fuel Pool Storage Racksi In this response, we desenbod our current programs which monitor Boraflex degradation in our spent fuel pool storage racks.
The results of our last blackness testing campaign which was performed August 1996 revealed gaps within the test area totaling more than six inches, which was the assumption used in the existing criticality analysis. Consequently, the test area was restricted from being loaded with freshly discharged fuel assemblies. The freshly discharged assemblies were loaded in racks outside of the test area. Loading areas outside of the test area would not pose an adverse condition due to the lower exposure which had been experienced by racks outside the test area. The test area had been exposed to freshly discharged fuel from six cycles. Thus, the test area Boiaflex configuration is conservative for the balance of the racks. However, this condition warrGnted a reanalysis of the test area's ability to maintain K.,5 0.95. The reanalysis was completed and issued to the plant on March 3,1997.
Following completion of the analysis, the loading of the test area was reevaluated and the loading restrictions were removed.
Results of Criticality Safety Analysis Th;s analysis addresses the storage of fuel assemblies with average enrichments up to 4.80 wt% in the central enriched zone. Top and bottom natural uranium blankets are explicitly included in fuel assembly modals. Credit for integral poison (Gadolinia) is included based on the peak reactiv!ty considering depletion effects throughout life. The analysis is based on a specific fuel mechanical type (GE1210x10), enrichment and Gadolinia configuration.
Other configurations are possible, and are also acceptable provided that they are less reactive than the base configuration. This base configuration was confirmed to bound both 8x8 and 9x0 configurations in use or stored at GGNS.
A conservative model was used which includes reactivity effects due to potential Boraflex degradation. This modelincludes gaps in the Botaflex due to shrinkage and an allowance for panel thinning associated with long term water ingress as described in IN 87 43 and GL 06-04. The analysis was performed in a 4 by 4 array of storage cells with repetitive boundary conditions. The Boraflex configuration for each panel was randomly selected from gap size, axiallocation and frequency distributions. Ten independent sets of rack configurations were analyzed and the results were statistically combined. The gap frequency distn%Ucq assumed that all gaps were located in the central 72 inches of the Botaflex panel and the totalloss due to Boraflex gaps ranged up to 12 inches. The panel loss assumptions along with the most recent measurement results are illustrated in Figure
- 1. This approach more realistically accounts for reactivity effects of Boraflex gaps while ushg very conservative gap assumptions.
The analysis was performed using the SCALE.4 system of analysis codes. The codes have been benchmarked against critical experiments to establish a bias and uncertainty. This benchmark was previously reviewed in support of the ANO 2 enrichment increase (Amendment No.178 to Facility Operating License No NPF 6 Arkansas Nuclear One, Unit No. 2, TAC No, M96478).
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Att:chrn:nt 1 to:
GNRO-97/00079 Page 2 of 2 The integral transport theory code, CASMO was used to perform depletion calculations to determine a reactivity credit for the presence of Gadolinia. This credit includes a 5 mk allowance for isotopic uncertainty. CASMO was also used to evaluate the uncertainties due to manufacturing tolerances Variations due to manufacturing are modeled as worst case values or treated as tolerances. Since all of the identified tolerances are independent, the reactivity for all tolerence parameters are statistically combined.
The calculated results include biases and uncertainties due to analytical methods and manufactunng tolerances in determining the 95/05 upper liniit k.a. The statistical treatment is equivalent to that used in the recent ANO 2 enrichment increase (Amendment No.178 to Facility Operating License No. NPF 0. Arkansas Nuclear One, Unit No 2. TAC No.
M90478). The analysis demonstrates that the GGNS high density fuel racks remain adequately subentical for the base fuel assembly and meet the requirements of NUREG-0800, and the GGNS Technical Specifications (i.e., k n s 0.95 under normal and abnormal conditions including all uncertainties). The upper limit k.n is determined to be 0.942, including all tolerances and uncertainties.
Figure 1 Panel Loi s Distribution Panet Loss Distributlen 0 50 0 45 - - - - -
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0 40 - - - - - - - ---
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0 35 0
0 30 0 25
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0 20 n
H 0 15 0 10 O c5 E "
0 00 1
2 3
4 5
6 7
8 9
10 11 12 Panel L.os s (in.)
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