LIC-06-0023, Response to Requests for Additional Information and Revision of Fort Calhoun Station Unit No. 1 License Amendment Request, Criticality Control During Spent Fuel Cask Loading in the Spent Fuel Pool

Response to Requests for Additional Information and Revision of Fort Calhoun Station Unit No. 1 License Amendment Request, Criticality Control During Spent Fuel Cask Loading in the Spent Fuel Pool

ML060870168
Person / Time
Site:	Fort Calhoun
Issue date:	03/27/2006
From:	Bannister D Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards
References
+sunsi/sispmjr=200604, LIC-06-0023, RIS-05-005
Download: ML060870168 (55)

Text

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Omaha Public Power District Fort Calhoun Station PO. Box 550, Highway 75 Fort Calhoun, NE 68023-0550 March 27,2006 LIC-06-0023 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

References:

1. Docket Nos. 50-285 and 72-054

2.

USNRC Regulatory Issue Summary 2005-05, "Regulatory Issues Regarding Criticality Analyses for Spent Fuel Pools and Independent Spent Fuel Storage Installations," March, 23, 2005 (ML043500532) (NRC-05-0038)

3.

Letter from OPPD (R. T. Ridenoure) to NRC (Document Control Desk), "Fort Calhoun Station Unit No. 1 License Amendment Request (LAR) 05-01 3, "Criticality Control During Spent Fuel Cask Loading in the Spent Fuel Pool,"

November 8,2005 (ML053120421) (LIC-05-0119)

4.

Teleconference between OPPD and NRC, Request for Additional Information for Fort Calhoun Station Unit No. 1 License Amendment Request, "Criticality Control During Spent Fuel Cask Loading in the Spent Fuel Pool," February 22,2006

SUBJECT:

Response to Requests for Additional Information and Revision of Fort Calhoun Station Unit No. 1 License Amendment Request, "Criticality Control During Spent Fuel Cask Loading in the Spent Fuel Pool9' In accordance with Reference 2, the Omaha Public Power District (OPPD) previously submitted the Reference 3 License Amendment Request (LAR) to address criticality control during spent fuel cask loading in the spent fuel pool. Attachment 1 of this letter provides OPPD's response to the Request for Additional Information questions discussed during the Reference 4 teleconference regarding the LAR. Attachment 2 of this letter provides clarifications to editorial items identified by the NRC Staff during their review of Reference 3.

No commitments to the NRC are made in this letter. I declare under penalt foregoing is true and correct. (Executed March 27, 2006)

y of pe jury that the Hi#hest in Customer Satislaction Wirn Resfdential Electric SemMce Amon# Medium-Sized UIilitias For the fifth year in a row, Omaha Public Power District has received a J.D.

Power and Associates award for Highest Customer Satisfaction With Residential Electric Service.

J.D. Power and Associates 2001 - 2005 Electric Utility Residential Customer Satisfaction Studiess! 2005 study based on a total of 26,782 consumer responses. The top 23 medium electric companies were ranked in the study. www.jdpower.com.

U.S. Nuclear Regulatory Commission LIC-06-0023 Page 2 If you have any questions or require additional information, please contact Mr. Thomas C.

Matthews at 402-533-6938.

Sincerely, D. J. ~ k i s t e r Plant Manager Fort Calhoun Station Attachments: 1) Response to Request for Additional Information

2) Clarifications to Omaha Public Power District's Letter LIC-05-0119

Enclosures:

1) NLTHOMS@-~~PT 50.68 Criticality Analysis for Fort Calhoun (Revision Number 1)

2) Revised Pages 2 and 5 of Attachment 1 of LIC-05-0119 and Revised 4.0-Page 2 of Clean-Typed Technical Specification cc: Director of Consumer Health Services, Department of Regulation and Licensure, Nebraska Health and Human Services, State of Nebraska

LIC-06-0023 Page 1 Omaha Public Power District (OPPD) Response to NRC's Request for Additional Information (RAI) on Fort Calhoun Station, Unit No. 1 License Amendment Request (LAR)

"Criticality Control During Spent Fuel Cask Loading in the Spent Fuel Pool" NRC Question 1:

The DSCposition in the spent fuel cavity is described as being 3 inches from the spent fuel racks.

Later in the application, the location is described as being 3 centimeters. State the correct spacing between the fuel racks and the cask and whether the model was conservative in accounting for neutron coupling in the criticality analysis.

OPPD Response to Question 1:

In the first paragraph of the License Amendment Request, Reference 3, section 4.1.3.6, Dry Shielded Canister, DSC, Position in the Spent Fuel Pool, SFP, Cavity, (page 12 of Attachment 1 to Reference 2) the unit of "inches" was incorrectly used. The fourth sentence should read, "In this orientation, the closest the transfer cask can approach the spent fuel racks is approximately three centimeters." The dimension of three centimeters is consistent with the dimension provided in section 6.5, DSC Position in the SFP Cavity, on page 23 of 36 of Enclosure 1 of Reference 2 and is the more conservative distance, i.e., the shortest distance, for neutron coupling.

NRC Question 2:

TS 4.3.2.2 Design Features states fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent may be stored in the new fuel storage racks. The 70.24 exemption states Ft.

Calhoun is licensed to a limit of 4.5 weight percent new fuel. Provide the LAR acceptance SE that allows Ft. Calhoun to store up to 5.0 weight percent in the new fuel storage racks.

Additionally, Sections 6.6 and 6.7 of Enclosure I of the LAR state a fresh fuel assembly of 4.55 weight percent enrichment is the most reactive fuel and this was the maximum enrichment offuel analyzed for the FCS spent fuel cask loading criticality analysis. Provide the technical justi$cation for not analyzing the most conservative accident condition where a fresh fuel assembly up to 5 weight percent is misloaded into the cask and k&I.

OPPD Response to Question 2:

Even though the new fuel storage racks were originally designed to store fuel with enrichment up to 5.0 weight percent, Technical Specification 4.3.1.1 restricts the spent fuel storage racks to fuel with no more than 4.5 weight percent enrichment. Based on this restriction and expected core

LIC-06-0023 Page 2 a maximum 4.5 weight percent enrichment. This procurement process is under the Fort Calhoun Station Quality Assurance Program, and ensures that no new fuel of greater than 4.5 weight percent enrichment could be received and accidentally loaded into the spent fuel cask. If future core designs include fuel with greater than 4.5 weight percent enrichment, OPPD will need to obtain NRC approval through appropriate licensing actions.

NRC Question 3:

Provide a table with the numerical values andparameters accounted for in the axial bias uncertainty applied to fuel burn-up greater than 30,000 MWD/MTU. How was +O. 01 3 Ak calculated and what burnup profile was used? (Farley used DOE axial burnup shape) Was this consistent with the burnup credit in the SFP criticality analysis previously performed?

OPPD Res~onse to Question 3:

The +0.013 Ak is consistent with the burnup credit in the SFP criticality analysis previously performed. The values and parameters of the Spent Fuel Racks Criticality Safety Analyses are presented in Tables 1 and 2 below. These tables were submitted as Attachment C of Reference C below in support of OPPD's application, made pursuant to 10 CFR 50.90, to increase the enrichment of the SFP racks approved by the NRC in Reference B. This request was subsequently approved by Reference D. The NRC staffs Safety Evaluation, given in Reference D, concluded that "Based on the review described above, the staff finds the criticality aspects of the proposed increased allowed enrichment limit of fuel stored in the Fort Calhoun spent fuel pool storage racks are acceptable and meet the requirements of General Design Criterion 62 for the prevention of criticality in fuel storage and handling."

The +0.013 Ak was taken directly from these OPPD licensing basis documents and is consistent with the burnup credit used in the SFP criticality analysis previously performed for the replacement SFP fuel storage racks. This axial bias is from HOLTEC reports HI-92828 (FC06025) and HI-95 1400 submitted in References A and C below and is therefore consistent with the burnup credit in the SFP criticality analysis previously performed under 10 CFR 50. As a result, the values and parameters consistent with the axial bias uncertainty pertaining to the 10 CFR Part 72 methodology, as requested by this question, were not re-calculated for that analysis and are not available.

The following Table 1 and Table 2 were submitted as Attachment C of Reference C in support of OPPD's application to increase the enrichment of the SFP racks.

LIC-06-0023 Page 3 Table 1 Summary of Criticality Safety Analyses Region I

~

~

Region 2 Design Basis 4.75% enrichment Temperature for analysis 4°C Reference k, (CASMO-3) 0.9392 Uncertainties In Bias B-10 loading Boral width Inner box dimension Water gap thickness SS thickness Fuel enrichment'"

Fuel densityu1 Eccentric position Statistical combination of uncenaintiesc' Burnup Uncertainty Axial Bumup Distribution

• 0.0024 i 0.003 1

• 0.0008

• 0.0009

• 0.0093

• 0.0004

• 0.0018

• 0.0022 Negative r 0.0106 Total 0.9392 = 0.0106 Maximum Reactivity (k) 0.9498

+ 0.0024

• 0.0032

• 0.0006

• 0.0011 N A 0.0002

• 0.0018 0.0022 Negative For fuel tolerances, uncertainties in Region 2 assumed to be the same as those for Region 1.

Square root of sum of squares.

LIC-06-0023 Page 4 Table 2 Evaluation of the 3linimum Burnup Requirements in Region 2 Initial Calculated Depletion Axial Burnup Limiting Enrichment kc, Uncert. A k Dist. A k Burnup From initial analysis

LIC-06-0023 Page 5

References:

A.

Letter from OPPD (W.G. Gates) to NRC (Document Control Desk), "Application for Amendment of Operating License," December 7, 1992, (LIC-92-304A)

B.

Letter from NRC to OPPD, "FCS Unit No. 1 Amendment No. 155 to Facility Operating License No. DPR-40," (TAC No. M85 1 16) August 12, 1993 (NRC-93-0292)

C.

Letter from OPPD (W.G. Gates) to NRC (Document Control Desk), "Application for Amendment of Operating License," February 1, 1996, (LIC-96-0010)

D.

Letter from NRC to OPPD, "FCS Unit No. 1 Amendment No. 174 to Facility Operating License No. DPR-40," (TAC No. M94789) July 30, 1996, (NRC-96-126)

LIC-06-0023 Page 6 NRC Question 4:

Provide the numerical value of the benchmark uncertainty obtained previously for the SFP criticality analysis at Ft. Calhoun to show the new benchmark uncertainty for the DSC criticality analysis is bounding and includes the uncertainties for the canister and the SFP. Provide a table that lists the tolerances, uncertainties and biases accounted for in the SFP/DSC criticality analysis and show these are bounding with respect to the existing SFP criticality analysis that credits burnup in the pool and kef6 0.95. Include tolerances such as eccentric positioning of fuel assemblies in storage cells since the spacing differs from the SFP racks.

OPPD Response to Question 4:

The benchmark uncertainty utilized in the DSC burnup credit criticality analysis is 0.01 549 in Ahff units. The benchmark uncertainty utilized in the spent fuel pool rack burnup credit criticality analysis is listed in HOLTEC report HI-95 1400 Table 1 (as noted in response to Question 3). The statistical combination of all uncertainty is given as 0.01 06 Ahff units for fuel in Region 1. A comparison of these two values indicates that the benchmark uncertainty utilized in the DSC burnup credit criticality analysis is conservative.

A table that lists the tolerances, uncertainties and biases with respect to the SFPIDSC criticality analysis has not been provided in this response because such a table does not exist in the criticality analysis. However, the underlying methodology in the construction of the design basis criticality analysis model renders such a table unnecessary. The design basis model is constructed utilizing the worst case geometric and material tolerances and results in the most conservative calculation of the system reactivity. A description of the methodology utilized to determine the design basis criticality model is given below:

The final hff value for the criticality analysis is expressed as follows:

kfinal = kkeno + 2akeno + Akmethod + Akgeometry + Akfuel < Safety Limits where kfinal is the final effective multiplication factor to be compared to the safety limits, kkeno is the calculated hff from the criticality safety code (KENO V.a),

Okeno is the uncertainty in kkeno, Akmethod is the bias due to the criticality analysis methodology and is the same as the benchmark uncertainty described above, AkgeomV is applied as a bias to the calculated kkeno based on the various geometricaVmateria1 tolerances pertaining to the Dry Shielded Canister -transfer cask, DSC I TC, and Akfuel is applied as a bias to the calculated kkeno due to the various geometricallmaterial tolerances pertaining to the representation of fuel in the criticality analysis model.

Note that this also includes the uncertainty in the fuel manufacturing, depletion parameters and fuel representation (axial burnup bias).

LIC-06-0023 Page 7 The Akgemet, is not explicitly calculated in the DSC criticality analysis. Rather, this is conservatively built into the criticality analysis model of the DSC 1 TC. The KENO model of the DSC I TC utilized in the criticality analysis is directly obtained from the bounding criticality analysis model (Chapter M.6, Criticality Evaluation for Standardized NLTHOMS~

FSAR, Revision 9, USNRC COC 72-1004) utilized in the part 72 criticality analysis. First, a series of evaluations that evaluate the reactivity effects of geometry and material tolerances of the various DSC I Cask components are performed. The design basis criticality model is then constructed by combining (not statistically combining) the worst reactivity effects of these parameters. For the NUHOMS~ -32PT DSC, the design basis criticality model is obtained by utilizing the nominal poison/aluminum plate thicknesses, minimum basket structure thickness, minimum fuel compartment width and minimum assembly to assembly pitch.

In addition, the most reactive placement of the fuel assemblies is evaluated in Table 4-6 (Page 18 of 36) of the criticality analysis, Enclosure of this letter. This evaluation also bounds the consideration of effects due to eccentric positioning of fuel assemblies.

In summary, the DSCITC criticality analysis model combines the worst reactivity effects of the various geometrylmaterial tolerances important to criticality while the SFP Rack criticality analysis results in a statistical combination of the uncertainties arising from the various geometrylmaterial tolerances important to criticality. Therefore, the treatment of geometry and material tolerances in the DSCITC analysis is bounding.

The Akbias is not explicitly calculated in the DSC criticality analysis. Rather, it is partly included in the CASMO model that provides the final burned fuel isotopic concentrations utilized in the criticality analysis and the remaining is included as an axial burnup bias (discussed in page 17 of 36).

Two sets of benchmarks are discussed in Page 16 of 36 - reference [8.12] that is a generic CASMO benchmark and reference [8.13] that specifically benchmarks the Fort Calhoun reactor operations for Cycles 17 through 20. The results of reference [8.13] benchmarks are briefly discussed in page 17 of 36 and demonstrate that CASMO calculations tend to overpredict the system reactivity for burnup credit. Since the CASMO predicted isotopic concentrations are directly input to the KENO V.a models, no additional bias due to manufacturing I operation is included in the final results. However, the axial burnup bias is included in the final keff value to overcome the non-conservatism due to the utilization of an axially uniform burnup profile within the fuel assembly.

LIC-06-0023 Page 8 NRC Question 5:

ClarzJfL the discrepancy between Summa y bullet I and page 21 of 36.

OPPD Response to Question 5:

The criticality analysis, Page 21 of 36, enclosed with LIC-05-0119 provided the cask assembly for a case with 500 ppm of soluble boron. The kegvalue listed did not include the 0.013 Aken:

This inadvertent omission is corrected below and the results continue to demonstrate the cask assembly kegis below 0.95.

I ~nrichment Bum-Up "1, U-235 (MWDIMTU) 1 3.5 24.110 The first bullet of the summary has been changed to reflect the correction discussed above and is now consistent Table 6-1 :

The most reactive DSC cask configuration (normal case) has a keg< 1 (i.e., 0.99713) with unborated water and a kefl < 0.95 (i.e., 0.92073) when flooded with borated water at a concentration of 500 ppm. Both cases apply burnup credit.

The correction above warranted further review of the report and a typographical error was found in the Table 6-1. Namely, the KENO kef+2a results for the 4.55 "1, U-235 case contained a typographical error which is corrected below. The correction is also applied to Figure 6-1.

Table 6-1 Maximum Enrichment and Burn-up Results for Type "A" and "B" Transfer Cask Enrichment "1, U-235 KENO kefl+2a Burn-Up (MWDIMTU) keg + 20 + Ake$

LIC-06-0023 Page 9 Bounding Keff Results SFP Peripheral Cells Empty 1 Acceptable 1 0 /

-r-r7-7-l o.97308 I

I I

I J 1.5 1.7 1.9 2.1 2.3 2.5 2.7 2.9 3.1 3.3 3.5 3.7 3.9 4.1 4.3 4.5 Fuel Assembly Enrichment wt-%

Figure 6-1 of this letter contains the revised Calculation Summary "NUHOMS-32PT 50.68 Criticality Analysis for Fort Calhoun," incorporating corrections as noted above.

NRC Question 6:

Clarzfv where the manufacturing, SFPparameters, and plant operation uncertainties are included in the calculations. Page 16 of 36 towards the bottom states they are not included as part of the benchmark. How were these values calculated? Include them on the table of Question 4.

OPPD Response to Question 6:

The clarifications for the location of the depletion related uncertainties are provided as part of the response to Question 4. However, the relevant portion is repeated here:

LIC-06-0023 Page 10 Two sets of benchmarks are discussed in Page 16 of 36 of the criticality analysis, Attachment 2 of this letter, - reference [8.12] that is a generic CASMO benchmark and reference [8.13] that specifically benchmarks the Fort Calhoun reactor operations for Cycles 17 through 20. The results of reference [8.13] benchmarks are briefly discussed in page 17 of 36 and demonstrate that CASMO calculations tend to overpredict the system reactivity for burnup credit. Since the CASMO predicted isotopic concentrations are directly input to the KENO V.a models, no additional bias due to manufacturing 1 operation is included in the final results.

The discussion regarding the inclusion of a Table containing the various uncertainties related to depletion is, therefore, not necessary since the benchmark calculations result in an over-prediction of kff, thereby rendering the CASMO predicted isotopic concentrations conservative.

NRC Question 7:

In the LAR, Ft. Calhoun states it included a 5 percent burnup uncertainty to the 4.55 enriched fuel on Table 6-1 of Enclosure I. There is no indication this uncertainty was also added to the less enriched fuel provided in Table 6-1. Section 5.A.5.d of the August 19, 1998, NRCguidance document, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," states the following: "A reactivity uncertainty due to uncertainty in the fuel depletion calculations should be developed and combined with other calculatingly uncertainties. In the absence of any other determination of the depletion uncertainty, an uncertainty equal to 5 percent of the reactivity decrement to the burnup of interest is an acceptable assumption. " Please provide the technical justz3cation why a 5 percent reactivity decrement to the burnup was not included in the calculations for burnup less than 30,000 MWD/MTU in accordance with the NRCguidance document referenced in the LAR.

Provide the numerical value ifa reactivity uncertainty due to uncertainty in the fuel depletion calculations was developed and combined with the other uncertainties for each enriched fuel modeled in the analysis and provided in Table 6-1.

OPPD Response to Question 7:

The reactivity decrement uncertainty due to burnup was incorporated by shifting the burnup curve upwards by 5% uniformly for all burnups and not just for the 4.55 weight percent enriched fuel as Table 6-1 of Enclosure 1 appears to indicate. (Note: Enclosure 1 of LIC-05-0119 has been updated and is included in this letter as Enclosure 1.) The text below the Table 6-1 has been modified to clarify application of this uncertainty to be consistent with the August 19, 1998, NRC guidance document.

LIC-06-0023 Page 1 Clarifications to Omaha Public Power District's Letter LIC-05-0119 During the NRC Staffs review of OPPD letter LIC-05-0119 (Reference 3 of this letter), three typographical items were questioned which are clarified as follows:

1. In Attachment 3, "Clean-Typed Technical Specification Pages," TS 4.3.1.3.b should state that "keff < 1.0 if fully flooded with unborated water,..." The " I

" sign should be a " < "

sign. Enclosure 2 of this letter contains the corrected page of the Clean-Typed Technical Specification Pages.

2. In Enclosure 1, "Framatome ANP Criticality Analysis," Section 6.1 should have read, "The CASMO results consist of the isotopic inventory used in KENO and these are documented in Reference 8.7 in Attachment A." The reference to "Reference 8.6 " was incorrect. This reference error has been corrected and other clarifications associated with the RAI have been made in the revised criticality analysis being submitted as the Enclosure 1 to this letter.

3. Throughout Attachment 1, the acronym of "DSC" should be referred to as "Dry Shielded Canister." The term "Dry Storage Canister," is synonymous to "Dry Shielded Canister,"

however, "Dry Shielded Canister" is the correct name which should be applied throughout Reference 3. Enclosure 2 of this letter contains corrected pages 2 and 5 (the only affected pages) of Attachment 1 of Reference 3.

LIC-06-0023 NUHOMS"-32~~

50.68 Criticality Analysis for Fort Calhoun (Revision Number 1)

Controlled Copy #2 Steve Andersen h

TRANSNUCLEAR Form 3.2-1 Calculation Cover Sheet Rev. No.:

AN ARBYA CoMPArn I

Calculation

Title:

I I

Project Name: NUHOMS" 32PT for OPPD NUHOMS@ -32PT 50.68 Criticality Analysis for Fort Calhoun Page:

I Software utilized: NIA since this is a documentation of editorial corrections I clarifications.

I Project No.:

DCR NO.:

Number of CDs attached: 0 If original issue, is I-lcenslng Review per TIP 3.5 required? Not Applicable since it is a Revision I Calculation is complete 1

1121 1121-008

~

No (explain)

Originator's Signature: Prakash Narayanan C-Date:03/24/2006 Project Engineer Signature:

of Yes Form 3.2-7, Revision I 37 Licensing Review No.

AN AREPA COMPANY I

Revision Surnrnarv I

Page: I 2A 1 of (

37 Rev. 0 Calculation 1

This calculation is prepared to incorporate OPPD comments and clarifications to the Framatome-ANP calculation, 86-9003453-000, "Fort Calhoun NUHOMS -

32PT Criticality Analysis." The TN file number for the Framatome-ANP calculation is 1121-0090. Framatome-ANP is an authorized supplierlvendor 1 under the Transnuclear QA (TIPS) program and therefore an independent review of the Framatome calculation is not performed here.

Rev. NO.:

The main body of this calculation is exactly the same as the Framatome ANP-calculation except for the changes as noted below. These changes would be incorporated with revision bars such that the nature of the change is clear and unambiguous.

1

1) Title of Table 6-1 changed to "Maximum Enrichment and Burn-up Results for Type " A and Type "B" Transfer Cask" in the List of Tables to be consistent with the actual table title.

2) Adjust the table (pqge 9) after Table 4-1 (before Section 4.2) such that the last row does not spill over to the next page.

3) Table 6-4 (page 26), added "OO" to the last entry in column 1 as it was missing.

I!

I I I

4) Section 7.0 (page 29), third bullet, 4.5 wlo was replaced with 4.55 wlo.

I I

5) Section 7.0 (page 29), added an additional bullet after the third bullet that describes a polynomial fit of the fuel burnup as a function of initial enrichment based on the results shown in Table 6-1 and Figure 6-5.

TRANSNUCLEAR AN AREVA COMPANY Calculation 1

Rev. No.: ~

1

~

1 Page: 1 28 of 1 37 Rev. 1 This calculation is revised to incorporate the changes as a result of a revision to the AREVNFANP calculation, 86-9003453. The current revision to the calculation is revision I. The revision was necessitated to provide responses to the NRC in order to correct typographical errors and clarify the treatment of burnup uncertainty in the criticality analysis.

The main body of the calculation is exactly the same as the AREVNFANP calculation except for the editorial enhancements detailed above for revision 0 of this calculation*. It may be noted that the total number of pages in the document is increased by 1 due to the inclusion of this revision summary. However, the revision summary page numbering has been changed to a different format so that the page numbering in the main document is consistent with that of the referenced AREVNFANP calculation. The footer section in the main document has been modified to include the total number of pages in the AREVNFANP document and the TN document for clarity.

For additional clarity, the description of the changes to the AREVNFANP document is detailed below. As in the previous revision, these changes would be incorporated with revision bars such that the nature of the change is clear and unambiguous.

1) A typographical error in Table 6-1 in page 21 has been corrected for the 4.55 wt. % U-235 case.

2) The note below Table 6-1 in page 21 has been replaced with language that specifies the burnup uncertainty applied.

3) A new note is added in Section 6.3, page 21 that describes the calculation of the kefi value and thus includes the burnup uncertainty in the calculated keW.

4) Figure 6-1 in page 22 has been updated to reflect the changes to the results shown in Table 6-1.

5) Update the summary in page 29 to reflect the changes to the results as shown in page

21. Specifically, the first bullet is modified to include these results.

In addition to these changes that are directly incorporated from the AREVNFANP calculation, the following are the changes that are identified by AREVNFANP and TN and are editorial in nature.

1) A typographical error in Section 6.1, Page 21 to replace "Reference 8.7 and in " with "Reference 8.xand in Attachment A."

63% 7

2) Correction of a typographical er the Revision 0 summary page of the calculation -

"86-90003453 to "86-9003453"

Transnuclear Calculation No. 1 121-0600 Rev. 1 86-9003453-00 1 TABLE OF CONTENTS LIST OF TABLES 4

LIST OF FIGURES...................................................................................................................

5 1.0 OBJECTIVE 6

2.0

SUMMARY

OF RESULTS 6

3.0 BACKGROUND

7 4.0 METHODOLOGY.........................................................................................................

7 4.1 CASMO-3 Calculations 8

4.2 KENO V.a Calculations 10 4.2.1 KENO V.a Benchmark Cases 10 4.2.2 Comparison of FANP SCALE versions 4.4a. 5.0. and 5.0.2 11 4.2.3 Changes to the TN KENO Model 11 4.2.3.1 Number of Histories 11 4.2.3.2 Removal of Soluble Boron and Gamma Shield....................................................... 12 4.2.4 Benchmark Uncertainties 14 4.2.5 Cask Manufacturing and Assembly Tolerances........................................................... 18 4.2.6 Fuel Assembly Position Evaluation............................................................................. 18 5.0 ASSUMPTIONS..........................................................................................................

20 5.1 Key Assumptions......................................................................................................... 20 5.2 Assumptions and Conservatisms.................................................................................

20 6.0 RESULTS.................................................................................................................... 21 6.1 CASMO Results...........................................................................................................

21 6.2 Normal Conditions KENO Results - Fresh Water....................................................... 21 6.3 Normal Conditions KENO Results - Borated Water................................................... 21 6.4 SFP Region 2 and DSC Type "A" Reactivity Comparison........................................ 22 6.5 DSC Position in the SFP Cavity.................................................................................. 23 6.6 Accident Condition - Misloaded Fuel......................................................................... 24 6.7 Accident Condition - Fuel Drop................................................................................... 26 6.8 DSC Type "A" and Type "B"......................................................................................

28 6.9 DSC Type "A" and Type "B" Loading Curve............................................................. 28 7.0

SUMMARY

ICONCLUSIONS.................................................................................... 29

8.0 REFERENCES

30 Attachment A: CASMO Isotopic Inventory for Burnup Credit 32 Framatome ANP Inc., an AREVA and Siemens Company Page 3 of 36. total pages = 37

Transnuclear Calculation No. 1 121-0600 Rev. 1 AREVAIFANP 86-9003453-001 LIST OF TABLES Table TITLE CE14x14 Fuel Dimensions for CASMO Runs Comparison of TN SCALE 4.4 results with FANP SCALE 4.4a Comparison of FANP KENO Output at 3.9 wlo Enrichment Comparison of FANP KENO Results for Different Histories 1.65 wlo Enrichment, Fresh Fuel, Pb and SS304 Shield System Bias Evaluation Results Maximum Enrichment and Bum-up Results for Type "A" and Type "B" Transfer Cask SFP Region 2 and DSC Reactivity Comparison Misloaded Fuel Assembly KENO kff Fuel Assembly Drop KENO keff Framatome ANP Inc., an AREVA and Siemens Company Page 4 of 36, total pages = 37

LIST OF FIGURES Figure TITLE Page K KENO VS. IMD for 0 ppm Boron at 1.65 wtlo U-235 14 Fuel Assembly Position Off-Set Inward - Model 19 Fuel Assembly Position Centered Case - Model 19 Fuel Assembly Position Off-set Outward - Model 19 Keno Bounding K-eff Results - SFP Peripheral Cells Empty 22 DSC and Transfer Cask Position in the SFP Cavity 23 Misloaded Fresh Fuel Bundle Locations 25 Dropped Fuel Bundle KENO Geometry 27 DSC Type "A" and "B" Loading Requirements 28 Framatome ANP Inc., an A R N A and Siemens Company Page 5 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAIFANP 86-9003453-001 1.0 OBJECTIVE The purpose of this report is to summarize the results from the criticality analysis that Framatome ANP (Advanced Nuclear Power) performed for loading operations with the Transnuclear, Inc. (TN) NUHOMSB-32PT Dry Shielded Canister (DSC) and transfer cask OS197L designs in the Fort Calhoun spent fuel pool. The license to operate the reactor includes a criticality safety evaluation of the spent fuel pool. The appropriate limits are noted in the plant Technical Specifications L8.'J. However, the Omaha Public Power District (OPPD) plans to improve the operation of Fort Calhoun with an Independent Spent Fuel Storage Installation (ISFSI). This ISFSI will incorporate Transnuclear Inc. (TN) NUHOMSB-32PT DSC to store the spent fuel. TN has an NRC Safety Evaluation Report (SER) to support the licensing of the NUHOMSB-32PT DSC in the Independent Spent Fuel Storage

~nstallation[~.~~.

While the Fort Calhoun Technical Specifications and the Transnuclear SER were considered sufficient licensin documents, in March of 2005 the NRC issued Regulatory Issue Summary (RIS) 2005-05 '83?

To comply with this RIS, the criticality safety analysis summarized in this report was required.

2.0

SUMMARY

OF RESULTS The TN criticality safety calculations of the NUHOMS@-32~~

DSC have been based on bounding "worst case" conditions [8.4, 8.51. That is, every parameter whose uncertainty could result in an increase in reactivity of the DSC was combined in the criticality model to represent the maximum increase. These bounding modeling conditions were carried over into the Framatome ANP (FANP) analysis of the DSC and transfer cask in the Fort Calhoun spent fuel pool. Moreover, the FANP calculations were benchmarked to previous TI1 results to show that FANP modeling would produce TN results for the cask criticality evaluations.

The existing Technical Specifications (TS) for the Fort Calhoun s ent fuel pool (SFP) allow burned uranium fuel in Region 2 as shown by TS Figure 2-10 [81y. FANP used a bounding modeling approach to duplicate the TS Figure 2-10 results. This benchmark demonstrated that the FANP modeling approach was consistent with the criticality safety analysis results that are the basis for the Technical Specifications. That is, using bounding conditions for every parameter whose uncertainty could result in a reactivity increase, and bounding the uncertainty bias from the model benchmarks, the FANP criticality safety analysis results confirmed the validity of the Fort Calhoun spent fuel pool Technical Specifications.

Consequently, the FANP modeling approach, with bounding parameters that produce the maximum kefl (effective neutron multiplication factor), is consistent with the licensing of the spent fuel pool, and the licensing of the TN NUHOMS@-32~~

DSC.

Framatome ANP Inc., an AREVA and Siemens Company Page 6 of 36, total pages = 37

Transnuclear Calculation No. 1 121-0600 Rev. 1 AREVAlFAN P 86-9003453-001 The FANP modeling of the NUHOMS@-32~~

DSC in the spent fuel pool with bounding conditions was used to perform the criticality safety analysis for the canister loading erati ti on[^.^]. The analysis considered normal and accident conditions. The normal condition analysis was performed with burnup credit in lieu of credit for the soluble boron in the spent fuel pool. The analysis determined the required fuel assembly bumup as a function of initial enrichment and equilibrium decay that maintains kg < 1 for the NUHOMS@-32~~

Transportable DSC assembly. The accident events considered were the misloading and dropping of a fresh fuel assembly of the highest enrichment permitted by facility Technical Specifications. The accident analysis determined the soluble boron concentration in the spent fuel pool of 800 ppm is required to maintain kef< 0.95 for the DSC assembly. All the cases apply bumup credit.

3.0 BACKGROUND

The Nuclear Regulatory Commission issued Regulatory Issue Summary (RIS) 2005-05 on March 23, 2005 regarding criticality analyses for spent fuel pools and independent spent fuel storage installation^[^.^'.

The NRC identified regulatory inconsistencies in licensee methodologies for criticality analyses and concluded that dry cask operations performed in the spent fuel pool (SFP) must meet both 10 CFR Part 72 and Part 50 requirements.

Transnuclear, Inc., an AREVA and Siemens Company, contracted with Framatome ANP, also an AREVA and Siemens company to perform the 10 CFR Part 50 criticality analyses for the NLTHOMS@-32~~

transportable dry shielded canister (DSC) and transfer cask OS197L in the OPPD Fort Calhoun SFP. The criticality analysis for the NUHOMS@-32~~

transportable DSC and transfer cask OS197L is documented in Framatome-ANP calculations 32-9003495-000 and 32-9001685-00 '8.71. This report summarizes the reference calculations.

4.0 METHODOLOGY The methodology applied in the criticality analysis for the NUHOMS'-32~~ transportable DSC and transfer cask OS197L documented in References 8.6 and 8.7 is the same methodolo y that was previously applied for the Fort Calhoun spent fuel pool criticality analysis f 891 and is based on the CASMO-3 and KENO V.a codes. This analysis is also consistent with previous Framatome-ANP analyses through the use of methods and benchmarks that have been previously reviewed and approved by the NRC for other utilities.

The most recent instances where the Framatome ANP methodologies have been submitted for review include:

Framatome ANP Inc., an AREVA and Siemens Company Page 7 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAIFANP 86-9003453-001 1

Date 1 PlanVDocket Accession ID Description Other previous submittals include the following:

ML052510504 ML052510502 USNRC Docket No. 50-305, "Kewaunee Fresh Storage and Spent Fuel Storage Pool".

USNRC Docket No. 50-346, "Davis Besse Fresh Storage and Spent Fuel Storage Pool".

USNRC Docket No. 50-302, "Crystal River 3 Spent Fuel Storage Pool" USNRC Docket No. 50-244. 'Ginna Spent Fuel Storage Pool."

The NUHOMS@-32~~

DSC assembly KENO V.a criticality analysis is documented in Reference 8.6. The CASMO-3 calculations are documented in Reference 8.7. The following sections of this report summarize these calculations.

0813 1/05 09/01/05 4.1 CASMO-3 Calculations The CASMO-3 (CASMO) calculations provided the isoto ic atom densities for the burned 14x14 fuel assembly used in the Fort Calhoun reactor The fuel enrichment evaluated ranged from 2.5 wt% U-235 to 4.75 wt%. The lumped fission product number densities in the CASMO depletion cases can not be used directly by the KEN0V.a code because they are not available in the KEN0V.a materials library. The lumped fission products 401 and 402 in CASMO have been appropriately modeled in the KEN0V.a calculation so that reactivity effects of the lumped fission products are preserved. The method used to convert the CASMO generated isotopic inventory to number densities for use in the KEN0V.a calculations was previously developed in Reference 8.8 and demonstrated to be conservative in Reference 8.7.

Shearon Harris/

50-400 The CASMO model for the Fort Calhoun fuel was developed using the fuel assembly geometrical information and sample code listings from Reference 8.8. This information is summarized in Table 4-1. The CE type 14x14 fuel assembly was loaded with 96% theoretical density fuel that was reduced to 10.3171 to account for dishing of the pellets. For conservatism, the fuel assemblies had no axial blankets. CASMO hot full power (HFP) depletions were performed and included the effects of non-uniform axial burnup as well as control rod insertion during operation. The depletion covered specific burnup points where the fuel isotopic inventory was needed for the subsequent KEN0V.a calculations.

Framatome-ANP Report 77-5069740-NP-00 License Amendment Request The fuel assemblies contained sixteen gadolinia bearing fuel rods with 4.0 wt% Gd for the 2.5 and 3.0 wt% enriched fuel. All the higher enrichment fuel assemblies contained sixteen Gadolinia bearing fuel rods with 8.0 wt% Gad rods.

Framatome ANP Inc., an AREVA and Siemens Company Page 8 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 86-9003453-001 Restart files were saved at the different burnup points where the isotopic inventory was to be calculated. The restart CASMO cases were calculated at the different burnup points of interest at cold conditions with zero ppm soluble boron and no control rods inserted. The control assembly and gadolinia assumptions conservatively overestimate the reactivity associated with operation in the spent fuel assembly.

Calculations were performed to effectively model the reactivity effects of the fission products.

Also, the short-lived isotopes were appropriately decayed.

Table 4-1 CE14x14 Fuel Dimensions for CASMO Runs Dimension Pellet Diameter Clad Inner Diameter Clad Outer Diameter Pitch GTLT Inner Diameter inches 0.377 0.384 0.440 GTLT Outer Diameter Arrav Width based upon pitch cm 0.95758 0.97540 1.11760 0.580 1.0350 Assembly pitch with Water Channel Total Fuel Length 1.4732 2.62890 1.1 150 8.12 Total Fuel Rods in Assembly 176 Temperature Information 2.83210 20.6248 8.18 128 HFP Moderator Temperature.

HFP Moderator Pressure HFP Boron 20.7772 325.12 Cold Moderator Temperature I Cold Fuel Temperature The cold restart case sets several of the short lived isotopes to zero to account for total decay using the CASMO multiplication option on the "CNU" card. The isotopes that are set to zero are the following:

566.35 OF 2100 psia p

p 500 Cold Moderator Pressure Cold Boron 570 OK HFP Fuel Temperature 144.79 bars 38.95 OF 38.95 OF Framatome ANP Inc., an AREVA and Siemens Company Page 9 of 36, total pages = 37 277 OK 277 OK 14.696 psia 0

Isotope 1077.53 OF 1.0 1325 bars CASMO ID 854 OK

Transnuclear Calculation No. 1 121-0600 Rev. 1 AREVAIFANP 86-9003453-00 1 4.2 KENO V.a Calculations 4.2.1 KENO V.a Benchmark Cases The initial KENO models for the T pe "A" DSC were provided by TN to FANP along with results from TN calculations [84, '! The FANP KENO is part of the SCALE version 4.4a code package operating on the Linux operating system platform. The TN KENO cases were from a SCALE 4.4 application on a Windows based PC platform. The differences between SCALE 4.4 and SCALE 4.4a have been previously documented in the open literature such as the SCALE newsletter.

Nevertheless, a benchmark exercise was performed in order to qualitatively assess the difference, if any, between the two KENO applications. The exact same cases were run at FAlVP and the resulting k@*

20 was compared to the TN kg* 20.

The minor differences observed is due to a combination of the differences in histories and the differences between SCALE version 4.4 and 4.4a. Results of the benchmark are presented in Table 4-2.

Table 4-2 Comparison of TN SCALE 4.4 results with FANP SCALE 4.4a Framatome ANP Inc., an AREVA and Siemens Company Page 10 of 36, total pages = 37

Transnuclear Calculation No. 1121-0600 Rev. 1 AREVAIFAN P 86-9003453-00 1 4.2.2 Comparison of FANP SCALE versions 4.4a, 5.0, and 5.0.2 The use of SCALE 4.4a was determined by a number of factors, including consistency with the previous calculation [8.81. The KENO calculation [861 showed that no significant change would have been introduced by the use of either SCALE version 5.0 or 5.0.2. A major difference in SCALE 5.0.2 is the code fix for an error related to cylindrical holes. The error may occur where the boundary of a cylindrical hole in a KENO model overlays the surrounding boundary. Because this is fixed in SCALE version 5.0.2, and there are no significant changes in results then it can be inferred that, even though holes are used in the TN model the usage does not encounter this error. It is also noted that no cylindrical holes occurred in the TN input files. The results from the comparison is presented in Table 4-3.

Table 4-3 Comparison of FANP KENO Output at 3.9 w/o Enrichment 4.2.3 Changes to the TN KENO Model The changes made to the KENO model involved the number of histories selected, the replacement of the lead shield in the transfer cask with stainless steel and removing the soluble boron. These changes are discussed in the following sections.

4.2.3.1 Number of Histories Illustrated in Table 4-4 are results with 500k, 1 OOOk, 2000k, and 4000k neutron histories. The aforementioned table shows that results are statistically equal for all cases, except for a couple of cases that slightly exceed 20. Therefore, cases with -1000k histories are sufficient to provide acceptable results. For conservatism, FANP elected to use 2500k histories (eg., gen

=2600, npg=1000, and nsk=100) in the continuing input decks.

Framatome ANP Inc., an AREVA and Siemens Company Page 11 of 36, total pages = 37

AREVNFANP Transnuclear Calculation No. 1 121 -0600 Rev. 1 86-9003453-001 Table 4-4 Comparison of FANP KENO Results for Different Histories 4.2.3.2 Removal of Soluble Boron and Gamma Shield The KENO criticality calculation utilized burnup credit in lieu of boron credit to satisfy the criticality safety criterion. This required that with no soluble boron the system remains subcritical (kff < 1). The TN KENO model was altered to remove the soluble boron and the interspersed moderator density (IMD) was adjusted from 5% to 100% to determine the most reactive case. All calculations used 2.5 million histories as previously discussed. Based on the Fort Calhoun Technical Specifications, a fresh fuel assembly with an enrichment of 1.65"/, was used as a starting point.

Framatome ANP Inc., an AREVA and Siemens Company Page 12 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAIFANP 86-9003453-00 1 The other modification to the model was the replacement of the lead (Pb) gamma shield in the transfer cask with stainless steel 304 (SS304) as presented by TN for the 'light' transfer cask OS197L [8.101. Table 4-5 summarizes the changes in results due to the change in shield materials. Figure 4-1 shows the KENO keflas a function of IMD values from 5% to 100% to determine the most reactive case.

Table 4-5 1.65 wlo Enrichment, Fresh Fuel, Pb and SS304 Shield Framatome ANP Inc., an AREVA and Siemens Company Page 13 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAIFANP 86-9003453-00 1 Figure 4-1 4.2.4 Benchmark Uncertainties kKeno vs IMD for Oppm Boron at 1.65 wt% 23SU 20%

40%

60%

80%

100%

IMD There are four independent benchmarks that FANP has used to establish the appropriate uncertainties for the criticality safety analysis of the Transnuclear NUHOMS@-32~~

DSC in the Fort Calhoun spent fuel pool.

(1)

The FANP KENO V.a calculations of the NUHOMS@-32~~

DSC were compared to those from Transnuclear. The Transnuclear calculations were independently benchmarked to a set of critical experiments; but more importantly, the Transnuclear calculations contained the worst case uncertainties to produce the maximum kg value.

(2)

The FANP KENO V.a calculations were used to model the Fort Calhoun spent fuel pool.

While the criticality safety modeling used to support the Technical Specifications incorporated an independent set of uncertainties, the FANP modeling should meet the licensing criteria for the Technical Specifications in order to demonstrate a consistent set of uncertainties.

(3)

The FANP KENO V.a calculations were used for benchmark comparisons to a set of critical measurements from cold experiments with the appropriate canister and spent fuel configurations. The cold experiments included plutonium buildup effects.

(4)

The FANP CASMO calculations were used for benchmark comparisons of the Fort Calhoun reactor operation during several reload cycles.

Framatome ANP Inc., an AREVA and Siemens Company Page 14 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAIFANP 86-9003453-00 1 The results from the benchmarks demonstrated the maximum reactivity that is associated with the uncertainties in the methods and models provide bounding results for the criticality safety modeling of the N U H O M S @ - 3 2 ~ ~

DSC. The results from the FANP benchmark comparisons to the Transnuclear KENO V.a calculations of the NUHOMS@-32~~

DSC demonstrated that the results were equivalent within the statistical uncertainty associated with 500,000 neutron histories.

The transition from the criticality safety modeling of the NUHOMS@-32~~

DSC in the Independent Spent Fuel Storage Installation to the loading of the DSC in the spent fuel pool required changing the boron concentration in the water as well as modeling the burned fuel.

The DSC in the Independent Spent Fuel Storage Installation had boron concentrations around 2000 ppm (parts per million boron).

However, the spent fuel pool criticality safety requirements are based on no boron. Therefore, the Transnuclear optimization of the highest reactivity conditions were repeated for the 0 ppm cases. The results followed those for the spent fuel pool. The highest density (one gram per cubic centimeter) was the most reactive.

This included both fresh and burned fuel. In addition, the assembly location in the canister with high boron concentrations continued to have the most reactive conditions with burned fuel and no boron. The other parameters that Transnuclear considered in the modeling of the canister, such as the minimum borated metal loading in the poison plates, remained optimized for the highest reactivity.

The results from the FANP benchmark comparisons with KENO V.a calculations to the Fort Calhoun Technical Specifications for the spent fuel pool demonstrated that the bounding modeling uncertainties produced the Region 2 Technical Specification results in Figure 2-10.[~.'] The bounding modeling uncertainties not only included the uncertainties associated with the spent fuel pool parameters, but also included the uncertainties associated with critical experiment benchmark comparisons and the modeling of plant operation and burnup for several cycles. Thus, while the FANP burned fuel modeling uncertainties in the spent fuel pool would not be expected to be equivalent to those in the safety analysis for the Technical Specifications, they do produce equivalent criticality safety results.

The FANP modeling approach, with bounding parameters that produce the maximum k@, is consequently consistent with the licensing of the spent fuel pool, and the licensing of the Transnuclear NUHOMS@-32~~

DSC. The KENO V.a results discussed in this document include the bounding uncertainties for the canister and spent fuel pool. The only additional uncertainties that need to be considered are those associated with the methods and those associated with the burned fuel isotopic concentrations.

An important part of the guidance that the IVRC and ANSI standards provide concerns the benchmark of the methods used for calculating kg.

The information stresses the importance of having the experimental conditions in the benchmarks essentially the same as those for the fuel and fuel cell models. Moreover, the standards note the requirement that the methods used to analyze the spent fuel pool models must be the same as those used to benchmark the experiments. The FANP criticality analysis to model the loading of the NUHOMS@-32~~

DSC in the spent fuel pool includes two sets of independent benchmarks in addition to the Framatome ANP Inc., an AREVA and Siemens Company Page 15 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAlFAhlP 86-9003453-00 1 CASMO benchmarks. One set of benchmarks includes experiments of specific configurations that are comparable to the Fort Calhoun spent fuel pool and the DSC.['"]

The second set includes experiments of fuel assembly configurations that are comparable to the Fort Calhoun fuel The KENO V.a benchmark calculations [' I modeled one hundred critical experiments that are representative of spent fuel pool and canister configurations. Twenty-one of the one hundred experiments were performed by FANP - B & W and are recommended by the NRC in its "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants" ['.14'.

This set included the effects of burnup by modeling plutonium fuel in addition to uranium fuel. While the standard recommends taking advantage of the statistical nature of the random uncertainties, the approach utilized for the NUHOMS@-32~~

DSC criticality safety analysis was to bound all uncertainties associated with the benchmark results. The bounding uncertainty is:

Bounding Benchmark Uncertainty = 0.01 549 Akef IVo calculation of the spent fuel pool - canister experiments gave a higher uncertainty than the above value.

Consequently, this value is bounding for subsequent criticality safety evaluations of loading the DSC in the Fort Calhoun spent fuel pool.

The second set of benchmarks is referenced in the "Reactor Analysis System for PWRs 9, [8.12]

report. The fuel design in the first set of benchmarks was not particularly representative of Fort Calhoun, but the second set specifically represented the fuel. Moreover, the first set validated the NITAWL -KENO methods while the second set validated the CASMO methods. The results of the calculations from the second set of benchmarks indicated that the bias and random uncertainties associated with CASMO were smaller than those associated with NITAWL - KENO. That is, no statistically significant bias could be observed and the random deviations were the result of the same type of parameters in the KENO V.a benchmarks. Consequently, no additional unique uncertainty was assigned to the CASMO methods with fuel assemblies of the Fort Calhoun type. Any kef that is predicted by CASMO includes the bounding uncertainties from the first set of benchmark results (0.01549 Ak@).

The FANP CASMO calculations were used for benchmark comparisons of the Fort Calhoun reactor operation during several reload cycles.

The uncertainties determined by this benchmark comparison are independent of the bounding conditions that are applied to the burned fuel based on the various tolerances for the components. Thus, the overloading of the uranium, replacing the grids with water, decreasing the burnup, increasing the axial burnup effects by modeling the insertion of control rods, overloading the burnable poisons, etc, are not included in this benchmark comparison.

The CASMO benchmark discussed in Reference 8.12 above was based on cold-clean critical experiments. However, the burnup effects of the fuel need to be benchmarked if burnup credit is to be utilized in the spent fuel pool loading of the canister. This set of benchmarks, in the "Ft. Calhoun PRISM Benchmarking Cycles 17-20" ['.I3' document, was used to assess CASMO burnup uncertainties. The calculations modeled the fuel in Cycles 17 through 20 Framatome ANP Inc., an AREVA and Siemens Company Page 16 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAlFAhlP 86-9003453-001 and followed the core operation throughout Cycles 17, 18 and 19. The results show a calculation bias that averages - 40 ppm boron at the beginning of cycles 17 through 19 and

- 26 ppm boron at the end. The calculations are more reactive; that is, the calculations require more boron to be critical than is measured. Thus, for applications to criticality safety the CASMO burnup results are considered to be approximately +.0025 Ap (delta-rho,

{delta-k)l k) too reactive. Consequently, they are conservative.

While the uncertainty associated with CASMO burnup benchmarks is conservative, there are uncertainty biases that are applied to ensure bounding criticality safety predictions. The dominant bias is from the axial burnup effects. The burnup credit calculations are performed assuming a uniform burnup profile throughout the active length of the fuel assemblies. The burnup profile for the burned fuel assemblies is not generally uniform due to the axial flux distribution in the core and the neutron leakage from the ends of the fuel assembly. This typically results in a burnup profile that looks resembles a "flattened cosine". The uniform burnup profile assumption results in the over-prediction of burnup at the ends of the fuel assembly and under-prediction of burnup in the fuel mid-region. The difference between the hff values based on the axial burnup profile and the uniform burnup assumption is what is termed as "axial end-effect" whose magnitude depends on the actual burnup value and the axial burnup profile selected. For the burnup credit calculations, the expected burnup values are such that the assumption of a uniform burnup profile may not be conservative. In other words, an axial end-effect bias needs to be applied to the burnup credit calculations to account for the increase in reactivity due to axial end-effects.

Generic analyses confirm the minor and generally negative reactivi effect of the axially distributed burn-up at values less than about 30,000 MWDIMTU[81671 AS a result, KENO calculations with less than 30,000 MWDNTU do not contain an axial bias. The highest burnup evaluated in this effort was 38,200 MWDIMTU. The axial bias uncertainty applied at this burn-up is +0.013 ~ k [ '. ' ~ ].

Another major contributor to the bounding uncertainty is the bias in the assembly burnup.

This bias results from inaccurate predictions of fuel assembly power, core power, and cycle lifetime. The modeling follows NRC guidance with a 5% uncertainty at the lower burnups expressed in terms of MWDIMTU (mega-Watt days per metric ton of uranium). When the fuel has been burned for several cycles, the bounding burnup is represented by 1565 MWDNTU, or an equivalent of 52 EFPDs (effective full power days) in Fort Calhoun.

While the application of these biases ensure conservative results with kfl values that are too high, the benchmark of the burned fuel to the Fort Calhoun spent fuel pool Technical Specifications shows that there is an overall consistency with the existing criticality safety analysis.

Because the approach taken with the criticality safety analysis for loading Transnuclear's NUHOMS@-32~~

DSC (canister) in the Fort Calhoun spent fuel pool is to treat uncertainties with bounding - biased values, the uncertainty from the benchmarks confirms that the uncertainties are appropriately bounding.

Framatome ANP Inc., an AREVA and Siemens Company Pqge 17 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAIFANP 86-9003453-00 1 The additional reactivity value that must be applied to the KENO V.a calculations of the loading model kg is the bounding bias from the benchmark comparisons. As noted above, this value is 0.0 1549 Ak@

4.2.5 Cask Manufacturing and Assembly Tolerances The bounding conditions for the N U H O M S @ - 3 2 ~ ~

DSC and fuel assemblies were determined by Transnuclear based on various tolerances for the components "" 8.S1. The most reactive system configuration was used for the present criticality evaluation.

However, the Transnuclear Part 72 evaluation was performed assuming soluble boron. Since the Part 50 criticality evaluation is performed with fresh water, the present work re-evaluated the moderator density and fuel assembly spacing assumptions to confirm the most reactive configuration was being used.

4.2.6 Fuel Assembly Position Evaluation The fuel assembly position evaluation for the Type "A" basket was performed by utilizing the most reactive Type A cask configuration and performing three cases; namely, off-set case, the centered case, and a symmetric offset case. The off-set configuration proved to be slightly more reactive. The models and results are shown on the following pages.

The most reactive system water density changed from an IMD of about 0.8 in soluble boron to an INlD of 1.0 in fresh water. The most reactive fuel assembly position is the same as in the soluble boron case. The position corresponds to an off-set toward the center of the cask.

These two conditions were combined to perform the criticality analysis. The results are summarized in Table 4-6.

Table 4-6 System Bias Evaluation Results Framatome ANP Inc., an AREVA and Siemens Company Centered with SS304 no Soluble Boron 4.75wt% at 38.96 GWDMTU Page 18 of 36, total pages = 37 IMD 80%

90%

100%

KENO 0.9365 1 0.95552 0.9691 1 SKENO 0.00039 0.00038 0.0004 1 TN Off-set with SS304 no Soluble Boron 4.75wt% at 38.96 GWDIMTU 80%

90%

100%

0.9391 3 0.95696 0.97105 0.00039 0.00039 0.00038 FANP Offset with SS304 no Soluble Boron 4.75 wt% at 38.96 GWDMTU 80%

90%

100%

0.93354 0.95 148 0.96522 0.0004 1 0.00038 0.00042

Transnuclear Calculation No. 1 121 -0600 Rev. 1 86-9003453-001 Figure 4-2 Fuel Assembly Position Off-Set Inward - Model Figure 4-3 Fuel Assembly Position Centered - Model Figure 4-4 Fuel Assembly Position Off-set Outward Model Framatome ANP Inc., an AREVA and Siemens Company Page 19 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAIFANP 86-9003453-001 5.0 ASSUMPTIONS 5.1 Key Assumptions A

1.--. - _ _ _ _ _ _ - - L : - - :- --_.

+:--

I:-:+-+:--

+I.,,+

h a.ra-:GaA --in-t n 11cl;nm tho the most reactive design.

NO PRA assemblies are modeled in the DSC.

The NUHOMS@-32~~

Type "A" DSC without PRA bounds the Type "B" DSC.

Fort Calhoun SFP peripheral cells adjacent to the Cask Pit Area are maintained empty during DSC loading.

No burnable poisons accounted for in any fuel assembly in the KENO model.

NO PRA assemblies are modeled in the DSC.

The transition rails between the basket and the canister shell is modeled as 100%

aluminum. Steel and open space in the transition rails reduces reactivity because these materials have much higher absorption cross-sections as compared to the aluminum.

All stainless steel is modeled as SS304. The small differences in the composition of the various stainless steels have no effect on results of the calculation CASMO cases assumed control rod was inserted for part of the depletion to maximize axial effects.

Water density was at the optimum moderator density of 1.OO gramlcc corresponding to 4°C.

All fuel rods are filled with fresh water in the pelletlclad gap for both normal and accident conditions.

All cases assume full DSC reflection in the radial direction and axially with a 20 cm reflector.

Only 90% credit is taken for the DSC B-10 in the poison plates.

Framatome ANP Inc., an AREVA and Siemens Company Page 20 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAIFANP 86-9003453-001 6.0 RESULTS 6.1 CASMO Results The CASMO results consist of the isotopic inventory used in KENO and these are documented in Reference 8.7 and in Attachment A.

I 6.2 Normal Conditions KENO Results - Fresh Water A series of CASMO / KENO runs has produced a curve to coincide in form with the technical specification of the Fort Calhoun Spent Fuel pool. However, the acceptance criteria are different; namely, the spent fuel pool acceptance criterion is for k,# < 0.95 whereas the curves in this report have an acceptance criterion of ke# < 1. It is understood that the Type "A" basket is limited to 3.9 W/O. Table 6-1 and Figure 6-1 summarize the results.

Table 6-1 Maximum Enrichment and Bum-up Results for Type "A" and "B" Transfer Cask The maximum ke#values in Table 6-1 include an axial bias uncertainty of 0.013 Ak when appropriate and a 5% burnup uncertainty for cases with fuel exposure.

6.3 Normal Conditions KENO Results - Borated Water The most reactive configuration from the fresh case was analyzed with 500 ppm of soluble boron. The results below demonstrate the cask assembly kef is below 0.95.

Framatome ANP Inc., an AREVA and Siemens Company Enrichment "lo U-235 3.5 Page 21 of 36, total pages = 37 I

Burn-Up (MWDIMTU) 24,110 KENO ke#+20 0.89224 kef + 20 + ~k~~~~

0.92073 1

I

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAIFAN P 86-9003453-001 6.4 SFP Region 2 and DSC Type "A" Reactivity Comparison The reactivity comparison between the DSC and SFP Region 2 was performed by selecting the DSC 3.50 "I, result from Reference 8.6 and the SFP Region 2 3.5 "I, case from Reference 8.16. The point is on Fort Calhoun Technical Specifications Figure 2-1'8.11. These results are summarized in Table 6-2.

Table 6-2 SFP Region 2 and DSC Reactivity Comparison Figure 6-1 Case SFP Region 2 DSC-Type A Bounding Keff Results SFP Peripheral Cells Empty I Acceptable I Enrichment "1, U-235 3.5 3.5 40,000 3 35,000 -

F 5 30,000 -

3 25,000 -

Q 3

E = 20,000 -

m >

5 E 15,000 -

S U) 5 10,000 -

S LL 5,000 -

0 1.5 1.7 1.9 2.1 2.3 2.5 2.7 2.9 3.1 3.3 3.5 3.7 3.9 4.1 4.3 4.5 Fuel Assembly Enrichment wt-%

Framatome ANP Inc., an AREVA and Siemens Company Burn-Up (MWDIMTU) 24,240 24,110 Page 22 of 36, total pages = 37 KENO kg 0.9148 0.9809

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVNFAN P 86-9003453-001 6.5 DSC Position in the SFP Cavity The DSC and transfer cask assembly is placed diagonally to the SFP fuel racks at approximately 45" due to physical limitations; that is, the cask can not fit in any other configuration. In the configuration shown in Figure 6-2, the closest the DSC transfer cask can approach the SFP rack is proximately 3 centimeters. That proximity is reached when the lifting hooks make physical contact with the SFP rack.

Figure 6-2 DSC and Transfer Cask Position in the SFP ~ a v i t ~ ' ~. ' ~ ~

,/

7 4 CLEARANCE WITH LIFTING HOOKS AT MAXIMUM ROTATION OF 4' The coupling between the SFP and the DSC was evaluated by assuming the SFP peripheral cells are maintained empty and reflective boundary condition applied to the DSC in the radial direction. The full reflection assumption is in effect a second DSC where the SFP Region 2 is located. This is a conservative approach since the DSC is more reactive than the SFP Region No. 2.

Frarnatome ANP Inc., an AREVA and Siemens Company Page 23 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 86-9003453-00 1 6.6 Accident Condition - Misloaded Fuel The misloaded fuel assembly accident analysis was performed by assuming the transfer cask is loaded with various enrichments as shown in Table 6-3 and the empty position is loaded with a fresh fuel assembly of 4.5 "I, enrichment. Multiple empty locations were assumed in order to locate the most reactive empty cell. The soluble boron concentration values ranged from 500 to 800 ppm. All analysis assumptions from the normal case were also applied to this analysis; namely, all fuel rod gaps are flooded with pure water and the cask is fully reflected. The most reactive fuel in this configuration is the 4.55 "I, enrichment. These results show the minimum required soluble boron concentration to maintain kfl

< 0.95 is 800 ppm with all uncertainties.

Table 6-3 Misloaded Fuel Assembly KENO k,ff See next page for the positions The analysis demonstrated the most reactive cells are toward the periphery of the cask array since these locations do not have boron plates. The misloaded positions are shown in Figure 6-3. The bounding value for the misloading fuel bundle accident at 800 ppm is the following:

Framatome ANP Inc.. an AREVA and Siemens Company Page 24 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAIFANP 86-9003453-001 CASE M1 Misload 4.5 w/o Fresh Fuel 0 cm of Water, Reflected CASE M2 Misload 4.5 w/o Fresh Fuel 0 cm of Water. Reflected CASE M3 Misload 4.5 w/o Fresh Fuel 0 cm of Water, Reflected CASE M4 Misload 4.5 w/o Fresh Fuel 0 cm of Water, Reflected Figure 6-3 Misloaded Fresh Fuel Bundle Locations Framatome ANP Inc., an AREVA and Siemens Company Page 25 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 86-9003453-001 6.7 Accident Condition - Fuel Drop The fuel drop accident was evaluated with the same fuel assembly soluble boron assumptions as in the misloading accident. The dropped assembly is placed along the perimeter of the transfer cask aligned longitudinally and evaluated at different azimuthal locations to find the most reactive position. Table 6-4 summarizes the results.

Table 6-4 Fuel Assembly Drop KENO keff DSC 3.5 "I, with Soluble Boron 500 pprn DSC 3.5 "1, with Soluble Boron 800 m m DSC 3.5 "I, with Soluble Boron 700 pprn kef+ 20 + Ak 0.95955 0.95899 0.94875 0.90793 Bundle Position '

0" 15" 30" 45" 0"

See next page for the positions DSC 3.9 "I, with Soluble Boron 800 pprn The 0" position is most reactive and the system bff is bounded by the misloading accident.

The bounding keffvalue with 800 pprn of soluble boron for the dropped fuel bundle accident is bounded by the misloading accident and is the following:

keff 0.94282 0.94246 0.92996 0.89 168 DSC 3.5 "1, with Soluble Boron 600 pprn 0.9 1023 0"

Framatome ANP Inc., an AREVA and Siemens Company CJ 0.00062 0.00052 0.00 165 0.00038 0"

Page 26 of 36, total pages = 37 0.00057 0.89554 0.92607 0.92686 DSC 4.55 "I, with Soluble Boron 800 pprn 0.00076 0"

0.00062 0.91255 0.94280 0.895 18 0.00066 0.92499

Transnuclear Calculation No. 1 121 -0600 Rev. 1 86-9003453-001 Dropped Assembly at 0' I

1 Dropped Assembly at 15' Figure 6-4 Dropped Fuel Bundle KENO Geometry Dropped Assembly at 30' Framatome ANP Inc., an AREVA and Siemens Company Dropped Assembly at 45' Page 27 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAIFANP 86-9003453-001 6.8 DSC Type "A" and Type "B" The reference KENO calculations were all performed in the most reactive DSC design. It corresponds to the Type "A" basket with no PRA and only 16 poison plates. The basket is limited in enrichment to 3.9 "I, by the licensing Certificate of Conformance (CoC)

The KENO calculations, however, were extended to 4.55 "I, and the results show the Type "A" DSC with 4.55 "I, meet the acceptance criterion. Therefore, any fuel cask up to 4.55 "I, with a minimum burn-up of 38,220 MWDMTU can be stored in any of the N U H O M S ~ ~ ~ P T DSC basket designs listed in the CoC.

6.9 DSC Type "A" and Type "B" Loading Curve The KENO results for the fuel enrichment and burn-up combinations listed in Table 6-1 support a loading curve for all the NUHOMS@~~PT DSC basket designs listed in the CoC.

Figure 6-5 is a composite curve for the Type A and B DSC basket designs.

Figure 6-5 DSC Type " A and Type "B" Loading Requirements SFP Peripheral Cells Empty Fuel Assembly Enrichment wt-%

Framatome ANP Inc., an AREVA and Siemens Company Page 28 of 36, total pages = 37

Transnuclear Calculation No. 11 21 -0600 Rev. 1 AREVAIFANP 86-9003453-001 7.0

SUMMARY

ICONCLUSIONS The DSC criticality analysis performed by FANP demonstrates the calculated loading curve complies with §50.68(4) as follows:

The most reactive DSC cask configuration (normal case) has a keff < 1 (i.e., 0.99713) with unborated water and a kef< 0.95 (i.e., 0.92073) when flooded with borated water I at a concentration of 500 ppm. Both cases apply burnup credit.

The bounding misloaded fuel assembly accident credits soluble boron and the results for 800 ppm is a keff value of 0.94035. Therefore, the minimum boron concentration required to maintain keff< 0.95 for accident conditions is 800 ppm with burnup credit.

Any fuel cask up to 4.55 "I, with a minimum bum-up of 38,220 MWDIMTU can be stored in any of the N U H O M S ~ ~ ~ P T DSC basket designs listed in the OPPD Certificate of Conformance.

The minimum required burnup as a function of initial enrichment can be expressed as a third order polynomial as shown below:

Burnup (MWDIMTU) = A + Bl*E + B ~

• E ~

+ B ~

• E ~

Where A

= -42324 B1

=36442 B2

=-7929.3 B3

= 837.1 E

= Initial Enrichment expressed as "I, U-235 The misloaded fuel bundle accident bounds the dropped assembly accident.

These results require spent fuel pool peripheral cells adjacent to the Cask Pit Area are maintained empty during DSC loading operation.

Framatome ANP Inc., an AREVA and Siemens Company Page 29 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAIFANP 86-9003453-00 1

8.0 REFERENCES

8.1 AREVAIFANP 38-9004068-000, "OPPD Tech Spec for SFP," 1 121 -01 00-23.

8.2 AREVAR ANP 38-9004083-000, TN Transmittal No. 1 121 -001 1-01, Amendment 8A, Chapter M6, Criticality Evaluation".

8.3 NRC Regulatory Issue Summary 2005-05, Regulatory Issues Regarding Criticality Analyses for Spent Fuel Pools and Independent Spent Fuel Storage Installations, March 23,2005.

8.4 AREVAJTN Doc. # NUH32PT.0600, "NUHOMS@-32~~

Transportable Dry Shielded Canister Criticality Analysis", Transnuclear West Calculational Package, 6/28/2001.

8.5 AREVAJTN Doc. # NUH32PT.060 1, "NUHOMS@-32~~

with 16 Poison Aluminum Plates", Transnuclear Calculational Package, approved 2/2/2004.

8.6 AREVAIFANP Calculation 32-9003495-000, NUHOMS-32PT Criticality Analysis for Fort Calhoun, October 2005.

8.7 AREVAJFANP Calculation 32-9001685-00, Fort Calhoun Transnuclear CASMO Analysis, October 18,2005 8.8 AREVARANP Calculation 32-5038209-00, Fort Calhoun SFP Criticality Analysis, February 26,2004.

8.9 AREVARANP Calculation 32-503877 1-00, Fort Calhoun Spent Fuel Pool CASMO

& Uncertainty Analysis, October 19, 2004.

8.10 AREVAIFANP 38-9004097-000, TN Transmittal No. 1 121 -00 1 1-0 1, Drawing Sketch 1 12 1-4035, "0s 197L Light Yoke Clearences in OPPD Spent Fuel Pool".

8.1 1 AREVAJFANP Doc. # 32-124532 1-00, "KENO V.a SCALE 4.2 Benchmark Calculations," L.A. Hassler, released 3/24/1997.

8.1 2 AREVAJFANP Doc. # EMF-96-029(P)(A), "Reactor Analysis System for PWRs,"

S.K. Merk, N.A. Anguiano, C.J. Lewis, R.W. Twitchell, A.H. O'Leary, R.G. Grummer, January, 1997.

8.13 AREVAJFANP Doc. # E-6088-NO 1-1, Revision 0, "Ft. Calhoun PRISM Benchmarking Cycles 17-20," K.C. Segard and N.A. Anguiano, signed off on 1/25/2002.

8.14 U.S. NRC Memorandum, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," from Laurence Kopp, August 19, 1998.

Framatome ANP Inc., an AREVA and Siemens Company Page 30 of 36, total pages = 37

Transnuclear Calculation No. 1 121-0600 Rev. 1 AREVAIFANP 8.15 AREVAEANP Doc. # 77-5069740, Revision 0, "Shearon Harris Criticality Evaluation" Appendix A - Bias and Bias Uncertainty Evaluation, August 2005.

8.16 AREVAEANP 32-9004446-000, "Design Input 1 12 1-0 100-26 EA-FC-96-00 1 Rev 0" Framatome ANP Inc.. an AREVA and Siemens Company Page 31 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAIFANP 86-9003453-001 Attachment A CASMO Isotopic Inventory for Burnup Credit The number densities for the various isotopes were determined from the running of CASMO in calculation 32-9001685-000 [8.71. While a significant number of CASMO and KENO runs were performed, the results presented in this attachment are the five sets values that are germane to the technical specification DSC loading curve. The isotopics were calculated in

[8.7] with special constraints regarding burnup and burnable poisons, in order to add conservatism to this calculation the burnable poison is not included as a part of the KENO input. The following tables present the isotopic number densities for the specific cases of interest.

Number Densities of Isotopes at 2.5 wlo Enrichment and 11.57 MWdIkg Burnup Isotope kr-83 rh-103 xe-131 CS-133 CS-134 CS-135 nd-143 nd-145 pm-147 sm-147 sm-149 sm-150 sm-151 sm-152 eu-153 eu-154 eu-155 U-234 U-235 U-236 U-238 np-237 pu-238 pu-239 PU-240 PU-24 1 pu-242 am-24 1 am-242m am-243 cm-242 cm-244 ag-109 Number Density 1.14978E-06 8.77403E-06 7.57771E-06 1.73247E-05 7.74486E-07 5.44835E-06 1.351 17E-05 5.60599E-05 4.06968E-06 6.006 16E-07 1.05617E-07 3.50666E-06 2.9913 1E-07 1.74630E-06 1.06085E-06 1.72340E-07 1.35399E-07 3.78897E-06 3.50755E-04 4.17397E-05 2.21288E-02 2.55799E-06 3.6308 1E-07 9.44123E-05 2.16590E-05 8.93673E-06 1.09121E-06 1.28453E-07 1.38667E-09 7.29087E-08 2.36401E-08 6.35366E-09 8.57360E-07 Framatome ANP Inc., an AREVA and Siemens Company Page 32 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAIFANP 86-9003453-001 Number Densities of Isotopes at 3.0 wlo Enrichment and 17.24 MWdIkg Burnup Isotope kr-83 rh-103 xe-131 CS-133 CS-134 CS-135 nd-143 nd-145 pm-147 sm-147 sm-149 sm-150 sm-151 sm-152 eu-153 eu-154 eu-155 U-234 U-235 U-236 U-238 np-237 pu-238 pu-239 pu-240 pu-24 1 pu-242 am-24 1 am-242m am-243 cm-242 cm-244 ag-109 Framatome ANP Inc., an AREVA and Siemens Company Number Density 1.64262E-06 Page 33 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVAIFANP 86-9003453-00 1 Number Densities of Isotopes at 3.5 wlo Enrichment and 22.90 MWdlkg Burnup Isotope kr-83 rh-103 xe-131 CS-133 CS-134 CS-135 nd-143 nd-145 pm-147 sm-147 sm-149 sm-150 sm-151 sm-152 eu-153 eu-154 eu-155 U-234 U-235 U-236 U-238 np-237 pu-238 pu-239 PU-240 pu-24 1 pu-242 am-241 am-242m am-243 cm-242 cm-244 ag-109 Framatome ANP Inc., an AREVA and Siemens Company Number Density 2.09699E-06 1.60 198E-05 1.35904E-05 3.2765 8E-05 2.39189E-06 1.33913E-05 2.49432E-05 1.08197E-04 5.95 153E-06 1.79783E-06 1.35956E-07 7.19843E-06 4.643 19E-07 3.07305E-06 2.53253E-06 5.77729E-07 3.29501E-07 4.54626E-06 3.7838OE-04 7.94066E-05 2.16453E-02 6.46629E-06 1.52726E-06 1.25355E-04 3.89142E-05 2.04652E-05 4.03443E-06 5.5 1327E-07 7.32830E-09 5.06057E-07 1.40253E-07 8.13660E-08 1.83072E-06 Page 34 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVNFAN P 86-9003453-001 Number Densities of Isotopes at 3.9 w/o Enrichment and 27.23 MWd/kg Burnup Isotope kr-83 rh-103 xe-131 CS-133 CS-134 CS-135 nd-143 nd-145 pm-147 sm-147 sm-149 sm-150 sm-151 sm-152 eu-153 eu-154 eu-155 U-234 U-235 U-236 U-238 np-237 pu-238 pu-239 pu-240 pu-24 1 pu-242 am-24 1 am-242m am-243 cm-242 cm-244 ag-109 Framatome ANP Inc., an AREVA and Siemens Company Number Density 2.41887E-06 1.82878E-05 1.54680E-05 3.82667E-05 3.18175E-06 1.66449E-05 2.88363E-05 1.27757E-04 6.3 1422E-06 2.34653E-06 1.42764E-07 8.55528E-06 5.14663E-07 3.45391E-06 3.1 1976E-06 7.82885E-07 4.25635E-07 4.8573 1E-06 3.93985E-04 9.2901 8E-05 2.15005E-02 8.1 5039E-06 2.3 1059E-06 1.27526E-04 4.21303E-05 2.39232E-05 5.73574E-06 7.80913E-07 1.10183E-08 9.00791E-07 2.23098E-07 1.8483 1E-07 2.17563E-06 Page 35 of 36, total pages = 37

Transnuclear Calculation No. 1 121 -0600 Rev. 1 AREVNFANP 86-9003453-001 Number Densities of Isotopes at 4.55 wlo Enrichment and 34.59 MWdIkg Burnup Isotope kr-83 rh-103 xe-131 CS-133 CS-134 CS-135 nd-143 nd-145 pm-147 sm-147 sm-149 sm-150 sm-151 sm-152 eu-153 eu-154 eu-155 U-234 U-235 U-236 11-238 np-237 PU-238 pu-239 pu-240 PU-24 1 pu-242 am-24 1 am-242m am-243 cm-242 cm-244 ag-109 Framatome ANP Inc., an AREVA and Siemens Company Number Density 3.0 1079E-06 Page 36 of 36, total pages = 37

LIC-06-0023 Revised Pages 2 and 5 of Attachment 1 of LIC-05-0119 Omaha Public Power District's Evaluation for Amendment of Operating License "Criticality Control During Spent Fuel Cask Loading in the Spent Fuel Pool" Revised 4.0-Page 2 of Clean-Typed Technical Specification Pages

LIC-0-0119 Page 2 1.0 DESCRPTION The Omaha Public Power District (OPPD) is requesting to amend Operating License DPR-40 for Fort Calhoun Station (FCS), Unit No. 1. The proposed changes would revise the FCS Technical Specifications (TS) to add limits and controls for spent fuel cask loading and unloading1 operations in the spent fuel pool. This License Amendment Request (LAR) is being submitted in response to NRC Regulatory Issue Summary (RIS) 2005-05 (Reference 7.1). An amendment to the FCS operating licensing is required to support dry shielded canister loading operations, scheduled to begin on or about March 1,2006.

Currently, the FCS TS include limits and controls for storage of unirradiated (fresh) fuel and spent fuel in the FCS spent fuel pool storage racks. Spent fuel cask loading in the spent fuel pool in support of dry shielded canister storage has not previously been performed at FCS. OPPD plans to implement dry spent fuel storage under the general license provisions of 10 CFR 72, Subpart K (Docket No.72-054) utilizing the Transnuclear Standard NUHOMS@ System (10 CFR 72 Certificate of Compliance No. 1004). As a result of our review of RIS 2005-05 and discussions with NRC staff, OPPD has determined that a Part 50 operating license amendment is necessary to support cask loading operations. Because the fuel basket inside the Transnuclear 32PT Dry Shielded Canister (DSC) has a different geometric spacing and neutron poison plate design than the FCS spent fuel storage racks, separate criticality analyses were required to demonstrate compliance with the Part 50 regulations, and corresponding new and revised Part 50 TS were deemed necessary.

Spent fuel cask loading will be performed in the northwest comer of the FCS spent fuel pool in a cask loading area adjacent to existing Region I1 spent fuel storage racks designated "D," "G2,"

and "E." The proposed TS changes are consistent with the assumptions and inputs used in the supporting criticality analysis (Enclosure 1).

The criticality analysis is consistent with previously accepted methodologies used in licensing actions for the FCS plant and at other nuclear power plants.

Recent similar license amendment requests submitted by Southern Nuclear Operating Company and Entergy Operations (References 7.2 and 7.3), and associated responses to NRC Requests for Additional Information ( M I ) (References 7.4 and 7.5) have been reviewed. OPPD has taken into consideration the content of those amendment applications and the issues discussed in the RAIs in developing this LAR, to the extent the information is applicable to FCS, in an attempt to reduce or eliminate any RAIs for this license amendment request.

The following sections include detailed information regarding the proposed changes, background, technical basis, regulatory requirements, no significant hazards, and environmental considerations associated with this license amendment request.

1 This LAR addresses any time the spent fuel cask is submerged in the spent fuel pool with one or more fuel assemblies in the cask during loading unloading operations. Hereafter in this LAR, only loading operations will be discussed for simplicity. The proposed TS are written appropriately to govern both loading and unloading operations.

LIC-0-0 1 19 Page 5 remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.

The spent fuel to be loaded into the NUHOMS@ System 32PT Dry Shielded Canister (DSC) at FCS cannot be shown by analysis to meet bff 5 0.95 with unborated water. Therefore, the second set of criteria from 10 CFR 50.68(b)(4) has been used to demonstrate compliance with 10 CFR 50.68(b)(l) for spent fuel cask loading at FCS for this LAR.

In order to demonstrate keff < 1.0 for the spent fuel when flooded with unborated water, the NRC Office of Nuclear Reactor Regulation (NRR) has historically permitted licensees to credit the reduced reactivity of the spent fuel associated with burnup during operation in the Part 50 criticality analysis. The NRC Spent Fuel Project Office (SFPO) has historically required a maximum value of bff (5 0.95) to be demonstrated with all fuel in the spent fuel cask assumed to be fresh fuel at the maximum enrichment allowed by the cask Certificate of Compliance (CoC) as described in NUREG-1536 (Reference 7.9). (Current SFPO review guidance does permit a limited amount of burnup credit to be considered.)

To date, no spent fuel storage systems have been licensed under 10 CFR 72 with burnup credit considered in the criticality analysis. Instead, the criticality analysts have taken credit for the negative reactivity of soluble boron in the spent fuel pool during loading operations for PWR fuel. Thus, Part 72 CoCs require soluble boron credit for certain PWR fuel storage systems to maintain spent fuel in the cask sufficiently subcritical during cask loading operations in the spent fuel pool. In addition, certain Part 50 criticality analyses also incorporate credit soluble boron in the spent fuel pool. However, the minimum soluble boron concentrations in the spent fuel pool required by the Part 50 and Part 72 Technical Specifications are also dependent upon differences in the storage system geometries and the amount of credit taken for neutron poison in the fixed neutron absorber in the spent fuel storage racks and spent fuel cask in the respective criticality analyses. These differences in criticality methodology and acceptance criteria, and the need to comply with both Part 50 and Part 72 during cask loading operations, are described in detail in RIS 2005-05.

FCS Dry Spent Fuel Storage As part of the long-term spent he1 management strategy at FCS, OPPD has decided to move some of its spent fuel assemblies currently in the spent fuel pool into dry storage at an on-site Independent Spent Fuel Storage Installation (ISFSI) under the general license provisions of 10 CFR 72, Subpart K. ISFSI operations are expected to begin in the first quarter of 2006 and proceed with periodic loading campaigns into the future. OPPD has chosen the Transnuclear Standard NUHOMS~ System using the 32PT DSC for dry spent fuel storage. OPPD will load the 32PT DSC under Amendment 8 to the CoC, which is expected to be effective on December 5, 2005 (FR Notice dated 9120105). Depending on the type of fuel basket in the 32PT DSC and the enrichment of the fuel to be stored, the Technical Specifications in the NUHOMS@ System 10 CFR 72 CoC require anywhere from 1800 to 2500 ppm soluble boron in the DSC for criticality control during wet loading of FCS spent fuel to preserve the assumptions made in the storage system Part 72 design basis criticality analyses.

4.0 DESIGN FEATURES (Continued)

c. A nominal 8.6 inch center to center distance between fuel assemblies placed in Region 2, the high density fuel storage racks,

d. A nominal 9.8 inches (East-West) by 10.3 inches (North South) center to center distances between fuel assemblies placed in Region 1, the low density fuel storage

racks,

e. New or partially spent fuel assemblies with a discharge burnup in the "acceptable domain" of Figure 2-10 for "Region 2 Unrestricted" may be allowed unrestricted storage in any of the Region 2 fuel storage racks in compliance with Reference (I),

f.

Partially spent fuel assemblies with a discharge burnup between the "acceptable domain" and "Peripheral Cells" of Figure 2-1 0 may be allowed unrestricted storage in the peripheral cells of the Region 2 fuel storage racks in compliance with Reference (1 1,

g. New or partially spent fuel assemblies with a discharge burnup in the "unacceptable domain" of Figure 2-10 will be stored in Region 1 in compliance with Reference (1).

4.3.1.2 The new fuel storage racks are designed and shall be maintained with:

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent,

b.

ken 5 0.95 if fully flooded with unborated water, which includes an allowance for

~~ncertainties as described in Reference (2).

c.

k,,

5 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in Reference (2),

d.

A nominal 16 inch center to center distance between fuel assemblies placed in the storage racks.

4.3.1.3 The spent fuel casks are designed and shall be maintained with:

a.

Fuel assemblies having a maximum U-235 enrichment of 4.5 weight percent,

b.

ken <1.0 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.5 of the USAR,

c.

keff 50.95 if fully flooded with borated water 2800 ppm, which includes an allowance for uncertainties as described in Section 9.5 of the USAR,

d.

A nominal 9.075-inch center-to-center distance between fuel assemblies placed in the spent fuel cask,

e.

Spent fuel assemblies with a combination of discharge burnup and initial average assembly enrichment in the "acceptable" range of Figure 2-1 1.

4.0 - Page 2 Amendment No.=