ML20087H623
| ML20087H623 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 04/26/1995 |
| From: | Tuckman M DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20087H627 | List: |
| References | |
| NUDOCS 9505040108 | |
| Download: ML20087H623 (19) | |
Text
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Duke ikwzr Company (704)3 73-4011 422 South Church Street Charlotte NC28242-0001 DUKEPOWER April 26,1995.
U.S. Nuclear Regulat:,ry Commission Attn: Document Control Desk Washington, D.C. 20555
Subject:
Catawba Nuclear Station. Units 1& 2 Docket Nos. 50-413 and 414 respectively Response to Request for Additional Information Proposed Fuel Enrichment Increase Gentlemen:
Enclosed, for your review, is Duke Power Company's response to your Request for Additional Information (R AI) dated January 4,1995. The RAI concerns proposed amendments to technical specifications, dated June 13,1994 for McGuire and Sept.19, 1994 for Catawba. The proposed amendments increase the allowable enrichment limit for fuel stored in our spent fuel pools. Responses to the questions applicable to Catawba Nuclear Station are enclosed in Attachment 1.
Accompanying changes to the specifications and technicaljustification are also enclosed in. These changes are based on specific requests in your RAI and subsequent reviews performed by site technical staff. They include: 1) changing the surveillance frequency for boron in the SFP's from once per 31 days to once per 7 days,2) removing the option to use alternate storage configurations in the SFP and replacing it with footnotes allowing specific analysis on attemate fuel types,3) changing the titles on Table 3.9-1 to better clarify qualifications for restricted and unrestricted storage,4) limiting the burnups in Table 3.9-2 to 60 GWD/MTU,5) adding information contained in the bases to the footnote of Figure 3.9-1,6) changing the bases to discuss the option to use specific analysis on alternate fuel and discuss the specific fuel design (s) used to develop the limits in Specification 3/4.9.13, and 7) changing specification 6.9 to require the SFP baron concentration limit to be specified in the COLR. Please replace the corresponding pages of the original submittal with these revised pages, as appropriate.
9505040108 950426 PDR ADOCK 05000413 P
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.o Additional changes to the amendment packages are also enclosed in Attachment 3. These changes modify the No Significant Hazards Analysis and Environmental Impact Statement for the amendment request based on conversations with your office and additional reviews by our staff. These changes provide additionaljustification for the amendment requests.
We appreciate your detailed review of our proposals and hope these responses are sufficient to satisfy your concerns. If you have additional questions or need additional information, please contact Ms. Judy Twiggs at 704-382-8897.
Sincerely, K. 3. TM M.S. Tuckman Senior Vice President Nuclear Generation jgt/ attachments U.S. NRC xc:
S.D. Ebeneter, Regional Administrator U.S. Nuclear Regulatory Commission - Region II 101 Marietta Street, NW - Suite 2900 Atlanta, Georgia 30323 R.E. Martin, Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 14H25, OWFN Washington, D.C. 20555 R.J. Freudenberger Senior Resident Inspector Catawba Nuclear Station i
Max Batavia, Chief Bureau of Radiological Health South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, S.C. 29201
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l M.S. Tuckman, being duly sworn, states that he is Senior Vice President of Duke Power Company; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this information conceming revisions to the Catawba Nuclear Station Facility Operating Licenses NPF-35 and NPF-52. He further acknolowdges that all the statements and matter set forth herein are true and correct to the best of his knowledge.
K. 3, M.S. Tuckman, Senior Vice President i
Subscribed and sworn to before me this Md -day of LA<2,1995.
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Notar Public
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My Commission Expires:
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ATTACHMENT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION CATAWBA PROPOSED FUEL ENRICHMENT INCREASE Ql) 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 average 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) liow do KENO Va calculations with the 123 group GMTH cross sections compare with CASMO-3/ SIMULATE-3 calculations for the same Catawba storage rack configuration?
A2) Calculated reactivities from both CASMO-3 and SIMULATE-3 were used to support this submittal as discussed in Section B.I. Comparisons of the calculated k-infinities between CASMO-3, SIMULATE-3 and KENO Va were performed for the Catawba spent fuel storage rack. The CASMO and SIMULATE k-infinities were noticeably higher than the KENO Va k-infinity. Table A2 below shows the results of this comparison.
Table A2 Fuel CASMO-3 SIMULATE-3 KENO Va Enrichment k-inf k-inf k-inf 4.1 0.93972 0.93972 0.91285 The relatively large differences between CASMO/ SIMULATE and KENO Va k-infinities are apparently due to CASMO over predicting reactivity for the large unpoisoned water gaps present in the Catawba spent fuel storage racks. Modeling sensitivity studies were performed in an attempt to reveal any possible modeling errors. No trends were found to indicate any problems with the models. Also, the CASMO code vendor was contacted. Discussions with the vendor concluded that th: models were accurate. Additional cases run by the vendor using different f
computer codes confirmed the tendency of this version of CASMO to overpredict reactivity for these geometries, l
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While the reactivity difference between CASMO/ SIMULATE and KENO Va are relatively large, the CASMO/ SIMULATE results are conservative. Since all the burnup credit criticality analysis for the spent fuel storage racks is based on the CASMO/ SIMULATE results, the proposed fuel storage requirements for the increased enrichment limit remain conservative.
Q3) Since DPC proposes 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 DPC's plans for a revision to the TS amendment application?
A3) You recently approved technical specification amendments 125 and 119 for Catawba to relocate several boron concentrations from the TSs to the COLR (SER dated October 7,1994). The spent fuel pool boron concentration from TS 4.7.13.3 was included in this list. Therefore, the list of NRC approved analytical methods in TS 6.9 does not need to be amended to include the methods for determining the spent fuel pool boron concentration. However, TS 6.9 does need to be amended to include the spent fuel pool baron concentration in the proposed TS 3/4.9.12.
included in this package are six new pages for Attachment I which include the revised TS 6.9.1.9 and new sample COLR pages. This information is also included in the new Attachment IV pages 8-12 through 8-24 which replace the current 8-12.
Q4) The NRC staff believes that the 31 day frequency for verifying spent fuel pool boron concentration stated in proposed SR 4.9.12 is too long, especially during fuel storage operations. We note that a comparable SR for ensuring subc..acality in the reactor during MODE 6 in the improved Westinghouse Standard TS is 7 days and that this is discussed in the BASES for those TS. We request that DPC provide further justification for the proposed SFP surveillance frequency. Any associated changes to the BASES should also be proposed including a discussion of the limiting SFP accident analysis A4) We concur that the surveillance requirement for verifying the Spent Fuel Pool (SFP) boron concentration in the Westinghouse STS is 7 days. The current surveillance interval for SFP boron is 31 days at Duke Power Company's McGuire Nuclear Station. Since McGuire is the only Duke Power facility currently requiring a spent i
fuel storage related TS, the current McGuire TS was used as the basis for the
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proposed Catawba TS.
The purpose of soluble boron in the SFP is to provide adequate criticality safety margin in the unlikely event of an accident which increases the reactivity of the pool.
Since the only pastulated accidents of this outcome involve the movement of fuel
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I assemblies, the SFP boron surveillance is linked with fuel movement in the SFP.
j Given that the minimum SFP boron concentration limits are currently 2475 ppm and l
that only about 500 ppm is needed to maintain k-eff below 0.95 in the event of an accident of this nature,it seems unlikely that a boron dilution of nearly 2000 ppm in 31 days over such a large volume of water could go undetected. However, TS 4.7.13.3.a requires that the SFP boron concentration be verified every 7 days.
Therefore, although the applicability for SRs 4.7.13.3.a and 4.9.12 differ, the surveillance interval in TS 4.9.12 will be changed from 31 days to 7 days to be consistent with TS 4.7.13.3.a.
included in this package are the proposed revisions to TS 3/4.9.12 in Attachment I and page 8-3 in Attachment IV of our original submittal.
Q5) The calculated worst case k-eff for the most reactive fuel assembly under optimum moderation conditions in the new fuel vault is given as 0.95861 and, therefore, meets the 0.98 criterion. What is the calculated worst-case k-eff for the fully flooded condition in the new fuel vault?
A5) The criticality analysis for the new fuel vault considered both fully flooded and optimum moderation conditions. Since, for the Catawba new fuel vaults, the optimum moderation condition is more limiting than the fully flooded condition, only the results of the optimum moderation condition are given in the submittal. A revision to page 7-1 of Attachment IV is provided in this package to include the results of the fully flooded condition. These results are summarized below.
The calculated worst case k-eff for the fully flooded condition in the new fuel vault is 0.93022, which meets the 0.95 criterion for the fully flooded condition.
Q6) DPCs proposed TS specifies that certain highly enriched fuel which could be used as filler assemblies would require assembly burnups of over 67 GWD/MTU M order to meet the NRC suberiticality requirements for spent fuel storage. However, the staff's current High Burnup Fuel Action Plan restricts burnups to currently approved l
levels (rod average of 60 GWD/MTU or less) because of recent experimental data which have shown a significant reduction of fuel failure thresholds for higher burnups.
Although the proposed TS relates only to spent fuel storage, the implication is that this burnup level is also acceptable in the reload core. As a result, we will not approve this aspect of your amendment request. We recommend that proposed TS Table 3.9-2 have an assembly cutoff which corresponds to the current rod average limit of 60 GWD/MTU.
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A6) As mentioned in Q6, the proposed TSs relate only to the storage of spent fuel and not the accumulation of burnup on the fuel, nor the associated limits placed on fuel burnup. The limits on maximum fuel burnup are verified for each reload core in the l
reload analysis performed for each cycle. Hence, the proposed TSs for spent fuel storage do not relate to the maximum fuel burnup limits. However, we understand your concern that with high burnup levels specified in the TSs, one could imply that it is acceptable to bum the fuel to these levels. Therefore, we have revised TS Table 3.9-2 to remove all burnups above 60 GWD/MTU. Included in this package are the revised pages for the:- TS Tables in Attachments I and IV.
Q7) We do not agree with proposed TS 3.9.13.c, which would allow fuel storage configurations other than those reviewed by the NRC, and request that it be deleted.
A7) We understand your concern for not wanting to allow fuel storage configurations other than those reviewed by your staff. The intent of this specification is to allow for specific criticality analyses for special situations without requiring additional TS changes. An example of this would be storage of fuel assembly designs not analyzed as part of this license amendment request, as a result of new fuel designs or shipments of fuel from another facility. Another, more likely, example would be storage of individual fuel pins as a result of fuel assembly reconstitution. A similar specification has been approved for McGuire Nuclear Station (March 24, 1987).
The specification was implemented at McGuire to accommodate storage of Oconee spent fuel shipped to McGuire for storage.
In response to your concern, we have revised TS 3.9-13. The changes include adding additional discussion in the BASES to reflect the intended use of this provision, adding a statement to Tables 3.9-1 and 3.9-2 indicating that specific analyses may be performed to qualify fuel assemblies for storage, and deleting proposed TS 3.9.13c. Included in this package are replacement pages for TS 3.9.13, TS Tables 3.9-1,3.9-2 and the BASES for TS 3/4.9.12 and 3/4.9.13 of Attachment I and pages 8-4,8-5,8-6 8-8 and 8-9 of Attachment IV.
Q8) As stated in Bases 3.9.12 and 3.9.13, the enrichments listed in Tables 3.9-1 and 3.9-2 are nominal enrichments and may exceed the listed value by a manufacturing tolerance of up to 0.05 weight percent U-235. Since the Bases are not a part of the TS, we suggest that the labels in these Tables be titled Initial Nominal Enrichment.
A8) We agree with the suggestion to include ' Nominal' in the titles for TS Tables 3.9-1 and 3.9-2 and believe that this provides needed clarity to this TS. New TS Tables are included in the replacement pages for Tables 3.9-1 and 3.9-2 of Attachment I and pages 8-5 and 8-6 of Attachment IV included in this package.
Q9) The Duke Power submittal for the McGuire proposed TS changes for fuel enrichment and storage, dated June 13,1994, states that the BWFC Mark BW fuel design is the most reactive of the three fuel types which exist at McGuire. The Duke
Power submittal for Catawba, dated September 19, 1994, states that the Westinghoose OFA design is the most reactive fuel of all fuel types stored at any Duke Power facility. Please discuss this apparent discrepancy.
A9) The most reactive fuel assembly design is dependent on the particular conditions.
Due to its wetter lattice, the OFA design is more reactive than the MkBW design for most moderated conditions at beginning of life (BOL). This includes the spent fuel pools and the new fud vaults with one exception. The MkBW design is more reactive under optimum moderation conditions in the McGuire new fuel vaults, due to subtle differences in the geometry. Also, due to its harder spectrum, the MkBW design becomes more reactive than the OFA design with depletion due to its increased plutonium production. The burnup at which the MkBW fuel becomes more reactive than the OFA increases with increasing enrichment.
All the analyses for the McGuire submittal used the MkBW design exclusively.
While the OFA design is more reactive for most BOL conditions, no fresh OFA assemblies are being used at McGuire. Furthermore, the inventory of all OFA assemblies was examined to ensure that all OFA assemblies are less reactive than MkBW assemblies of the same enrichment and burnup by verifying that sufficient burnup exists on theses OFA assemblies. Therefore, the use of only the MkBW design for McGuire is justified. If a fuel assembly design other than the MkBW design is to be used at McGuire in the future, an analysis wil'. be performed to verify that this design is bounded by previous analyses, or a new TS amendment package will be submitted to define new burnup versus enrichment limits.
For Catawba, the spent fuel pool analysis explicitly modeled the BWFC MkBW fuel design, and the Westinghouse standard (STD) and OFA designs. The most reactive of these three designs, for a given burnup and enrichment, was used to set the final TS limit. The Oconee fuel designs were compared to the above designs to verify that the Oconee fuel is less reactive at BOL conditions. However, this comparison did not consider two future Oconee fuel designs which are more reactive than all cuirent designs at BOL. Therefore, the option of storing Oconee fuel at Catawba is rescinded and will be handled on a case by case basis, if needed. Appropriate language is incorporated in the revised page 5-1 of Attachment IV included in this package.
The new fuel vault analysis for Catawba used the OFA design exclusively, which is the most reactive design for this storage configuration for both the fully flooded and optimum moderation conditions.
In conclusion, the statements concerning the most reactive fuel designs in the McGuire and Catawba subm;ttals are accurate. These statements are in the context of new fuel vault calculations. The MkBW fuel design is the most reactive of the three fuel types which exist at McGuire, since no fresh OFA assemblies are stored, or planned for storage at McGuire. For the Catawba new fuel vault, the OFA design
i is the most reactive fuel of all fuel types currently stored at any Duke Power facility, excluding future Oconee fuel designs which are planned. Finally, although the
.MkBW fuel design is more reactive than the OFA design for burned fuel, the criticality analysis explicitly modeled all 17x17 designs for the Catawba spent fuel pool.
Q10)Section VII.3 discussing region interface restrictions appears to be incomplete.
Please supply the missing information.
A10) Two additional lines of text for Section VII.3 were missing from the top of page 7-3.
This information is included in the replacement for page 7-3 included in this package.
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ATTACHMENT 2 REVISED P!;
This attachment includes the revisions to Attachments I and IV of the Catawba Proposed Technical Specification Changes originally submitted on September 19,1994 l
ATTACHMENT I PROPOSED TECHNICAL SPECIFICATION CHANGES This section contains the proposed modifications to the CNS Technical Specifications.
In general, these changes increase the initial fuel enrichment limit and establish restricted loading patterns, and associated burnup criteria, for qualifying fuel in the Catawba Spent Fuel Pools. These changes are necessary to improve core reload designs and increase operational flexibility, while at the same time maintaining acceptable criticality safety margin and decay heat removal capabilities. In addition, several administrative changes have been included in order to provide clarity to the Specifications and bring them more in line with STS format. A description of each of the changes being requested is given below.
The accompanying FSAR changes will be incorporated at the next annual revision l
following approval of this submittal. These changes are identified and discussed in Section Vill of Attachment IV.
- 1. The Technical Specification Index is being changed to add Specifications 3/4.9.12 and 3/4.9.13, accompanying Tables 3.9-1 and 3.9-2, and Figure 3.9-1. This change is purely administrative in nature.
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- 2. Specification 3/4.9.12, Spent Fuel Pool (SFP) Boron Concentration is being added to establish an explicit SFP boron concentration limit, where one does not exist, as well i
as to establish a consistent LCO for SFP boron concentration for all modes operation.
By allowing this limit to be specified in the COLR, the change reduces the possibility of a dilution event since all other potential sources of borated water to the SFP are also specified in the COLR. It also provides consistency with other operational, cycle specific limits. This change is being made in order to accommodate the more complex SFP storage requirements and provide clarity to the Specifications, as well as increase I
operational flexibility. The change also provides more consistency with STS format.
- 3. Specification 3/4.9.13, accompanying Tables 3.9-1 and 3.9-2, and Figure 3.9-1 are being added to establish restricted loading patterns (with appropriate interface restrictions) for spent fuel storage and associated burnup criteria. The proposed changes are necessary to increase the efficiency of fuel storage while at the same time ensuring that acceptable criticality safety margin and decay heat removal capabilities are maintained. The format of these changes is also more in line with STS format. The technical basis for these changes and the associated criticality analysis are described in detailin Attachment IV.
- 4. The BASES for Sections 3/4.9.12 and 3/4.9.13 of the Technical Specifications has been added to reflect the addition of the corresponding Specifications and to more fully explain the basis for each LCO, Action Statement and Surveillance Requirement covered by these Specifications. Paragraph 2 of the BASES explains the provisions to use specific analysis for fuel types not previously analyzed, in addition the specific fuel designs used to develop the limits in Specification 3/4.9.13 are discussed. Paragraph 3 of the BASES explains the acceptability of using less reactive fuel components or non-fuel components in designated fuel assembly locations and non-fuel components in empty celllocations, as this would ensure the reactivity limits are met while increasing operational flexibility. In addition, the last paragraph specifies the limit for maximum fuel Revision i 1
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enrichment,5.00 weight %, as the basis for all fuel storage requirements imposed by Technical Specification 3/4.9.13 and to describe appropriate methods for interpolating the data provided in Tables 3.9-1 to 3.9-5. The proposed modifications to the BASES Section are also more consistent with those in STS.
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- 5. Technical Specification 5.6, Fuel Storage, has been changed to reflect appropriate limits for criticality analysis for fuel storage. This change allows increased operational flexibility, while maintaining acceptable criticality safety margin. In addition the Specifications have been reformatted to bring them more in line with STS format.
- a. Specification 5.6.1 has been changed to allow for use of kett s 0.98 under optimum moderation conditions in the rack design criteria for new fuel storage racks. Actual calculations have shown that kett s 0.95, under all storage conditions however, this change allows increased flexibility when performing criticality analyses and is consistent with the criteria currently specified in ANSI-ANS57.3,1983 and STS.
- 6. Specification 6.9 has been changed to require specification of the SFP boron concentration limit in the Core Operating Limits Report (COLR). This change is being made to ensure consistency with corresponding changes to Specifcation 3/4.9.12 and is administrative in nature.
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REFUELING OPERATIONS 3/4.9.12 SPENT FUEL POOL BORON CONCENTRATION 1
LIMITING CONDITION FOR OPERATION l
3.9.12 The boron concentration in the spent fuel pool shall be within the limit specified in the COLR.
APPLICABILITY:
1 During storage of fuelin the spent fuel pool.
l ACTION:
- a. Immediately suspend movement of fuel assemblies in the spent fuel pool and initiate action to restore the spent fuel pool boron concentration to within J
its limit.
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- b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS:
y 01a h n
EM 4.9.12.
Verify at least once per 7 days that the spent fuel pool boron concentration 3r
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is within its limit.
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Revision 1
3/4.9.13 SPENT FUEL ASSEMBLY STORAGE LIMITING CONDITION FOR OPERATION I
3.9.13 New or irradiated fuel may be stored in the Spent Fuel Pool in accordance with l
these limits:
- a. Unrestricted storage of fuel meeting the criteria of Table 3.9-1; or l
- b. Restricted storage in accordance with Figure 3.9-1, of fuel which does nat dj meet the criteria of Table 3.9-1.
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I APPLICABILITY:
During storage of fuel in the spent fuel pool.
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ACTION:
- a. Immediately initiate action to move the noncomplying fuel assembly to the correct location.
- b. The provisions of Specification 3.0.3 are not applicable.
1 SURVEILLANCE REQUIREMENTS:
4.9.13 Prior to storing a fuel assembly in the spent fuel storage pool, verify by I
administrative means the initial enrichment and bumup of the fuel assembly j
are in accordance with Specification 3.9.13.
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I Table 3.9-1 L
L Minimum Qualifyina_ Burnuo Versus Initial Enrichment for Unrestricted Storaos 1
Initial Nominal Enrichment Assembly Burnup (Welaht% U-235)
(GWD/MTU) 4.05 (or less) 0 4.50 2.73 5.00 5.67 M
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UNRESTRICTED E
STORAGE 6?gg}'
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$2 RESTRICTED SI 4
STORAGE g
tigStlY a
st#T 4.00 4.25 4.50 4.75 5.00 gd/M.h <<
Initial Nominal Enrichment (Weight % U-235)
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Fuel which differs from those designs used to determine the requirements of Table ddd
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3.9-1 may be qualified for Unrestricted storage by means of an analysis using NRC v
approved methodology to assure that k, is less than or equal to 0.95.
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3/f,7<lg[C)l Likewise, previously unanalyzed fuel up to 5.0 weight % U-235 may be qualified for Restricted storage by means of an analysis using NRC approved methodology to assure that k, is less than or equal to 0.95.
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Revision I
l Table 3.9-2 Minimum Qualifyina Burnuo Versus initial Enrichment
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for Filler Assemblies h) a MJ l
Initial Nominal Enrichment Assembly Burnup s
,, g v#,n (Weicht% U-235)
(GWD/MTU) 1.90 (or less) 0 2.00 16.83 2.50 26.05 3.00 35.11 hlb 3.50 43.48 4t cpfd p
4.00 51.99 j
op 4.48 60.00 g (o 70 UNACCEPTABLE l
For Use As Filler Assembly g
60 oUf l
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@ 50 ACCEPTABLE q 40 For Use As Filler Assembly Eg 30 j' 20 UNACCEPTABLE g
For Use As Filler Assembly
.12 10 Mel 1.50 2.00 2.50 3.00 3.50 4.00 4.50 u
lnitial Nominal Enrichment (Weight % U-235)
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- g4ha, Fuel which differs from those designs used to determine the requirements of Table nNd 3.9-2 may be qualified for use as a Filler Assembly by means of an analysis using g
NRC approved methodology to assure that k, is less than or equal to 0.95.
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Revision 1 l
Fiaure 3.9-1 Reauired 3 out of 4 Loadina Pattern for Restricted Storace RESTRICTED RESTRICTED RESTRICTED RESTRICTED FUEL FUEL FUEL FUEL d
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RESTRICTED FILLER RESTRICTED FILLER FUEL LOCATION FUEL LOCATION v
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w y
vy RESTRICTED RESTRICTED RESTRICTED RESTRICTED FUEL FUEL FUEL FUEL e-i RESTRICTED FILLER RESTRICTED FILLER l
FUEL LOCATION FUEL LOCATION dtE l guf Restricted Fuel:
Fuel defined for Restricted Storage in Table 3.9-1. (Fuel defined 40 for Unrestricted Storage in Table 3.9-1, or non-fuel components, Cg/,fy or an empty location may be placed in restricted fuel locations as needed)
Filler Location:
Either fuel which meets the minimum burnup requirements of Table 3.9-2, or an empty cell.
Boundary Condition:
Any row bounded by an Unrestricted Storage Area shall contain a combination of restricted fuel assemblies and filler locations arranged such that no restricted fuel assemblies are adjacent to each other.
Example: In the figure above, row 1 or column 1 can Dol be adjacent to an Unrestricted Storage Area, but row 4 or column 4 can be.
Revision 1
BASES 3/4.9.12 and 3/4.9.13 SPENT FUEL POOL BORON CONCENTRATION and SPENT FUEL ASSEMBLY STORAGE The requirements for spent fuel pool boron concentration specified in Specification 3.9.12 ensure that a minimum boron concentration is maintained in the pool. The requirements for spent fuel assembly storage specified in Specification 3.9.13 ensure that the pool remains suberitical. The water in the spent fuel storage pool normally contains soluble boron, which results in large suberiticality margins under actual operating conditions. However, the NRC guidelines based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting k n Of e
l 0.95 be evaluated in the absence of soluble boron. Hence the design of the spent fuel storage racks is based on the use of unborated water, which maintains the spent fuel pool in a suberitical condition during normal operation with the pool fully loaded. The double contingency principle discussed in ANSI N-16.i-1975 and the April 1978 NRC letter (Ref. 4) allows credit for soluble boron undes other abnormal or accident conditions, since only a single accident nee.1 be considered at one time. For example, the most severe accident scenario is assoc ded with the accidental misloading of a fuel assembly. This could increase the re&ty of the spent fuel pool. To mitigate this postulated criticality related accident, boron is dissolved in the pool water.
Tables 3.9-1 and 3.9-2 allow for specific criticality analyses for fuel which does not meet the requirements for storage defined in these tables. These analyses would require using NRC approved methodology to ensure that k n s 0.95 with a 95 percent o
probability at a 95 percent confidence level as described in Section 9.1 of the FSAR.
This option is intended to be used for fuel not included in previous criticality analyses.
4 dl Fuel storage is still limited to the configurations defined in TS 3.9-13. The use of specific analyses for qualification of previously unanalyzed fuel includes, but is not dd limited to, fuel assembly designs not previously analyzed which may be as a result of new fuel designs or fuel shipments from another facility. Currently analyzed fuel designs include the Babcock and Wilcox MkBW design, and the Westinghouse Standard and Optimized fuel designs. Another more likely, and expected use of this specific analysis provision would be to analyze movement and storage of individual fuel pins as a result of reconstitution activities.
In verifying the design criteria of k n s 0.95, the criticality analysis assumed the most o
conservative conditions, i.e. fuel of the maximum permissible reactivity for a given configuration.
Since the data presented in Specification 3.9.13.a and 3.9.13.b represents the maximum reactivity requirements for acceptable storage, substitutions of less reactive components would also meet the k n s 0.95 critoria. Hence an empty cell, o
or a non-fuel component may be substituted for any designated fuel assembly location.
These, or other substitutions which will decrease the reactivity of a particular storage cell will only decrease the overall reactivity of the spent fuel storage pool.
If both restricted and unrestricted storage is used, an additional criteria has been imposed to ensure that the boundary row between these two configurations would not locally increase the reactivity above the required limit.
Revision 1
The action statement applicable to fuel storage in the spent fuel pool requires that action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress.
This is most efficiently achieved by immediately suspending the movement of fuel assemblies. Prior to the resumption of fuel movement, the requirements of the LCOs must be met. This requires restoring the soluble boron concentration and the correct fuel storage configuration to within the corresponding limits. This does not preclude movement of a fuel assembly to a safe position.
The surveillance requirements ensure that the requirements of the two LCOs are satisfied, namely boron concentration and fuel placement. The boron concentration in the spent fuel pool is verified to be greater than or equal to the minimum limit. The fuel assemblies are verified to meet the suberiticality requirement by meeting either the initial enrichment and burnup requirements of Table 3.9-1 and 3.9-2, or by using NRC approved methodology to ensure that kg s 0.95.
By meeting either of these requirements, the analyzed accidents are fully addressed.
The fuel storage requirements and restrictions discussed here and applied in specification 3.9.13 are based on a maximum allowable fuel enrichment of 5.0 weight %
U-235.1he enrichments listed in Tables 3.9-1 and 3.9-2 are nominal enrichments and include uncertainties to account for the tolerance on the as built enrichment. Hence the as built enrichments may exceed the enrichments listed in the tables by up to 0.05 weight % U-235. Qualifying burnups for enrichments not listed in the tables may be linearly interpolated between the enrichments provided. This is because the reactivity of an assembly varies linearly for small ranges of enricament.
REFERENCES
- 1. " Regulatory Guide 1.13: Spent Fuel Storage Facility Design Basis", U.S. Nuclear Regulatory Commission, Office of Standards Development, Revision 1, December 1976.
2 " Design Objectives for L;ght Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations", American Nuclear Society, ANSI N210-1976/ANS-57.2, April 1976.
- 3. FSAR, Section 9.1.
- 4. Double contingency principle of ANSI N16.1-1975, as specified in the April 14,1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).
Revision 1
.