ML061000606
| ML061000606 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 04/10/2006 |
| From: | Wang A Plant Licensing Branch III-2 |
| To: | Ridenoure R Omaha Public Power District |
| Wang A | |
| Shared Package | |
| ML061000597 | List: |
| References | |
| TAC MC8860 | |
| Download: ML061000606 (17) | |
Text
April 10, 2006Mr. R. T. Ridenoure Vice President - Chief Nuclear Officer Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.
Post Office Box 550 Fort Calhoun, NE 68023-0550
SUBJECT:
FORT CALHOUN STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT RE:(TAC NO. MC8860)
Dear Mr. Ridenoure:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosedAmendment No. 239 to Renewed Facility Operating License No. DPR-40 for the FortCalhoun Station, Unit No. 1. The amendment consists of changes to the Technical Specifications (TSs) in response to Omaha Public Power District's (OPPD's) application dated November 8, 2005, as supplemented by letters dated March 17 and 27, 2006. In the March 17,2006, letter, OPPD committed to be in full compliance with Section 50.68 of Title 10 of the Codeof Federal Regulations 6 months after the fall 2006 refueling outage is complete.The amendment adds limits and controls for the spent fuel cask loading and unloadingoperations in the spent fuel pool (SFP). The change modifies the TSs by adding a new Limiting Condition for Operation (LCO) 2.8.3(6) that establishes (1) a boron concentration requirementduring cask loading operations in the SFP, and (2) a spent fuel burnup-initial enrichment limit in the spent fuel cask to ensure subcritical conditions are maintained during spent fuel cask loading operations in the SFP. In addition, the change modifies TS Tables 3-4 and 3-5, and adds a new subsection 4.3.1.3 in Design Features 4.3.1 to describe the spent fuel cask designfeatures. The change also conforms pagination.A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will beincluded in the Commission's next biweekly Federal Register notice.Sincerely,/RA/Alan B. Wang, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-285
Enclosures:
- 1. Amendment No. 239 to DPR-402. Safety Evaluationcc w/encls: See next page April 10, 2006Mr. R. T. Ridenoure Vice President - Chief Nuclear Officer Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.
Post Office Box 550 Fort Calhoun, NE 68023-0550
SUBJECT:
FORT CALHOUN STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT RE:(TAC NO. MC8860)
Dear Mr. Ridenoure:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosedAmendment No. 239 to Renewed Facility Operating License No. DPR-40 for the Fort Cal hounStation, Unit No. 1. The amendment consists of changes to the Technical Specifications (TSs) in response to Omaha Public Power District's (OPPD's) application dated November 8, 2005, as supplemented by letters dated March 17 and 27, 2006. In the March 17, 2006, letter, OPPDcommitted to be in full compliance with Section 50.68 of Title 10 of the Code of FederalRegulations 6 months after the fall 2006 refueling outage is complete.The amendment adds limits and controls for the spent fuel cask loading and unloadingoperations in the spent fuel pool (SFP). The change modifies the TSs by adding a new Limiting Condition for Operation (LCO) 2.8.3(6) that establishes (1) a boron concentration requirementduring cask loading operations in the SFP, and (2) a spent fuel burnup-initial enrichment limit in the spent fuel cask to ensure subcritical conditions are maintained during spent fuel cask loading operations in the SFP. In addition, the change modifies TS Tables 3-4 and 3-5, and adds a new subsection 4.3.1.3 in Design Features 4.3.1 to describe the spent fuel cask designfeatures. The change also conforms pagination.A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will beincluded in the Commission's next biweekly Federal Register notice.Sincerely,/RA/Alan B. Wang, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-285DISTRIBUTION
- PUBLICGHill (2)
Enclosures:
- 1. Amendment No. 239 to DPR-40LPLIV ReadingRidsNrrDorlDpr2. Safety EvaluationRidsNrrDorl (CHaney/CHolden)RidsNrrDorlLplg (DTerao)cc w/encls: See next pageRidsNrrPMAWangRidsNrrLALFeizollahi RidsOgcRp RidsAcrsAcnwMailCenter RidsRegion4MailCenter (DGraves)YHsiiRidsNrrDirsItsbACCESSION NO.: (Pkg) (TS) OFFICENRR/LPL4/PMNRR/LPL4/LANRR/CSGB/BCOGCNRR/LPL4/BCNAMEAWangLFeizollahiJNakoskiSUttal (NLO)DTerao DATE4/10/064/10/063/30/064/7/064/10/06DOCUMENT NAME: E:\Filenet\ML061000606.wpdOFFICIAL RECORD COPY OMAHA PUBLIC POWER DISTRICTDOCKET NO. 50-285FORT CALHOUN STATION, UNIT NO. 1AMENDMENT TO RENEWED FACILITY OPERATING LICENSEAmendment No. 239License No. DPR-401.The Nuclear Regulatory Commission (the Commission) has found that:A.The application for amendment by the Omaha Public Power District (thelicensee) dated November 8, 2005, as supplemented by letters dated March 17 and 27, 2006, complies with the standards and requirements of the AtomicEnergy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;B.The facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission;C.There is reasonable assurance (i) that the activities authorized by thisamendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with theCommission's regulations;D.The issuance of this license amendment will not be inimical to the commondefense and security or to the health and safety of the public; and E.The issuance of this amendment is in accordance with 10 CFR Part 51 of theCommission's regulations and all applicable requirements have been satisfied. 2.Accordingly, Renewed Facility Operating License No. DPR-40 is amended by changesto the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No. DPR-40 is herebyamended to read as follows:B.Technical SpecificationsThe Technical Specifications contained in Appendix A, as revised throughAmendment No. 239 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.3.The license amendment is effective as of its date of issuance and shall be implementedwithin 30 days of issuance.FOR THE NUCLEAR REGULATORY COMMISSION/RA/David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical SpecificationsDate of Issuance: April 10, 2006 ATTACHMENT TO LICENSE AMENDMENT NO. 239 RENEWED FACILITY OPERATING LICENSE NO. DPR-40DOCKET NO. 50-285Replace the following pages of the Appendix A Technical Specifications with the attachedrevised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.REMOVEINSERTTOC - Page 8TOC - Page 82.8 - Page 142.8 - Page 14 2.8 - Page 152.8 - Page 15 2.8 - Page 162.8 - Page 16 2.8 - Page 172.8 - Page 17 2.8 - Page 182.8 - Page 18 2.8 - Page 192.8 - Page 19 2.8 - Page 202.8 - Page 20 2.8 - Page 212.8 - Page 21 2.8 - Page 222.8 - Page 22 2.8 - Page 232.8 - Page 23
2.8 - P age 24 ------------2.8 - P age 25 ------------2.8 - P age 26 ------------2.8 - P age 273.2 - Page 53.2 - Page 5 3.2 - Page 123.2 - Page 12 4.0 - Page 24.0 - Page 2 4.0 - Page 34.0 - Page 3 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONRELATED TO AMENDMENT NO. 239 TO RENEWED FACILITYOPERATING LICENSE NO. DPR-40OMAHA PUBLIC POWER DISTRICTFORT CALHOUN STATION, UNIT NO. 1DOCKET NO. 50-28
51.0INTRODUCTION
In March 2005, the Nuclear Regulatory Commission (NRC) issued Regulatory Issue Summary(RIS) 2005-05, "Regulatory Issues Regarding Criticality Analyses for Spent Fuel Pools and Independent Spent Fuel Storage Installations," to (1) alert addressees at pressurized-water reactor (PWR) facilities to findings suggesting that the spent fuel pool (SFP) licensing anddesign bases and applicable regulatory requirements may not be met during loading, unloading, and handling of dry casks in the SFPs; (2) emphasize the importance of maintaining subcriticalconditions for spent fuel storage in moderated environments; and (3) encourage addressees to review the current SFP and independent spent fuel storage installation (ISFSI) licensing and design bases at their facilities to ensure compliance with regulations during dry cask loading,unloading, and handling operations. As a result of the RIS, Omaha Public Power District (OPPD/the licensee) submitted anapplication dated November 8, 2005 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML060930053), as supplemented by letters dated March 17 and 27, 2006 (ADAMS Accession Nos. ML060790081 and ML060870168, respectively),
requesting changes to the Technical Specifications (Appendix A to Renewed Facility OperatingLicense No. DPR-40) for the Fort Calhoun Station, Unit No. 1 (FCS). In the March 17, 2006, letter, OPPD committed to be in full compliance with Section 50.68 of Title 10 of the Code ofFederal Regulations (10 CFR) 6 months after the fall 2006 refueling outage is completed.The proposed amendment would revise FCS Technical Specifications (TSs) to add limits andcontrols for the spent fuel cask loading and unloading operations in the SFP to ensure its continued compliance with NRC regulations governing the safe handling of irradiated fuel in theSFP. Specifically, the proposed changes would revise the FCS TSs by adding a new Limiting Condition for Operation (LCO) 2.8.3(6) that establishes (1) an SFP boron concentrationrequirement during cask loading operations in the SFP, and (2) a spent fuel burnup-initial enrichment limit in the spent fuel cask to ensure subcritical conditions are maintained during spent fuel cask loading operations in the SFP. In addition, the proposed amendment wouldalso modify TS Tables 3-4 and 3-5, and add a new subsection 4.3.1.3 in Design Features 4.3.1to describe the spent fuel cask design features, and conform pagination. The FCS TSs currently permit the licensee to store unirradiated (fresh) fuel and spent fuelassemblies in the FCS SFP storage racks. However, OPPD plans to implement dry caskstorage in an ISFSI facility at FCS in accordance with the general license provisions of 10 CFRPart 72, Subpart K, using the Transnuclear (TN) Standard NUHOMS System. OPPD intendsto load spent fuel into the NUHOMS-32PT dry shielded canister (DSC) in its spent fuel caskloading area for subsequent removal and dry storage on the ISFSI. The supplemental letters dated March 17 and 27, 2006, provided additional information thatclarified the application, did not expand the scope of the application as originally noticed, anddid not change the staff's original proposed no significant hazards consideration determinationas published in the Federal Register on December 20, 2005 (70 FR 75494).
2.0REGULATORY EVALUATION
Because the fuel basket inside the NUHOMS
-32PT DSC has a different geometry spacing andneutron poison plate design than the FCS spent fuel storage racks, separate criticality analysesare required to demonstrate compliance with the Part 50 regulations.While in the SFP for wet loading operation, both Part 50 and Part 72 requirements pertaining tocriticality control apply to spent fuel cask loading operations. General Design Criterion (GDC) 62 in Appendix A to 10 CFR 50 specifies that criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometricallysafe configurations. The acceptance criteria specified in 10 CFR 50.68 section (b)(1) and (b)(4) for criticalityprevention in the SFP that are applicable to the licensee's proposed amendment are,respectively: 1.Plant procedures shall prohibit the handling and storage at any one time of more fuelassemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water; and 2.The effective neutron multiplication factor (Keff) of the spent fuel storage racks loadedwith fuel of the maximum fuel assembly reactivity, including an allowance for uncertainties, must not exceed 0.95 at a 95 percent probability at a 95 percentconfidence (95/95) level, if flooded with borated water, and the Keff must not exceed 1.0(subcritical) at a 95/95 level, if flooded with unborated water.
Under 10 CFR 72.124, "Criteria for nuclear criticality safety," the NRC regulates dry caskstorage activities to ensure that subcriticality is maintained during the handling, packaging, transfer, and storage of spent fuel assemblies. The NRC regulations for dry cask criticalityprevention rely on favorable geometric configurations and fixed neutron absorbers. However, unlike 10 CFR 50.68, the 10 CFR Part 72 regulations for criticality prevention in dry casks allowlicensees to credit the SFP soluble boron for maintaining subcritical conditions during cask loading, unloading, and handling operations in the SFP. Therefore, many cask designs haveincorporated soluble boron credit, in lieu, of a burnup credit as a means of increasing day cask storage capacity while maintaining subcritical conditions. OPPD's amendment request proposes to demonstrate that it can satisfy the applicable 10 CFR 50.68 criticality prevention requirements, with a burnup credit, during cask loading, unloading, and handling operations in the SFP. The NRC defines the acceptable methodologies for performing criticality analyses in thefollowing documents: 1.NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports forNuclear Power Plants," Section 9.1.2, Draft Revision 4, "Spent Fuel Storage;" 2.Proposed Revision 2 to Regulatory Guide 1.13, "Spent Fuel Storage Facility DesignBasis;" and 3.NRC Memorandum from L. Kopp to T. Collins, "Guidance on the RegulatoryRequirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," August 19, 1998.
The NRC staff used these documents to assist in its review of the licensee's amendmentrequest to ensure compliance with GDC 62 and 10 CFR 50.68.
3.0TECHNICAL EVALUATION
3.1Spent Fuel Cask System DescriptionThe TN Standard NUHOMS System incorporates a number of DSCs for dry storage of spentfuel. Each DSC includes a variety of options for the fuel basket design to offer operational flexibility. Different fuel basket designs may be used based on the characteristics of the fuel to be loaded and the level of SFP soluble boron desired. OPPD chose to use the NUHOMS-32PT PWR DSC for dry storage of the FCS spent fuel. NUHOMS-32PT DSC hasthree types of fuel baskets, namely, Type A, B, and C, which are differentiated by the maximum permitted fuel enrichment, neutron poison plate configuration, number of poison rod assemblies (PRA), and soluble boron. The cask user may choose 16, 20, or 24 poison plates in the fuel basket based on the enrichment of the fuel to be loaded and the desired soluble boron limit.
The PRAs are inserts required by the Standard NUHOMS Part 72 Certificate of Compliance. The PRAs are installed in certain fuel storage locations in the NUHOMS-32PT DSC fuelbasket designs for criticality control. There are 0, 4, and 8 PRAs in the Type A, B, and C baskets, respectively. Operationally, an empty DSC with an integral fuel basket is inserted into a TN OS197L transfercask (TC) and the assemblage is placed in the cask loading area of the FCS SFP. Except for the floor of the cask loading area being approximately two feet below the floor of the SFPproper, there is no physical barrier between the cask loading area and the spent fuel racks. Upto 32 fuel assemblies, meeting the specified limits, can be moved from the fuel storage racks into the NUHOMS
-32PT DSC. Upon completion of fuel movement, while st ill under water ashield plug is installed into the NUHOMS
-32PT DSC. The DSC/TC assemblage is thenremoved from the SFP for completion of DSC preparation for deployment at the ISFSI.Table 3.0-1 in the Framatome Advanced Nuclear Power (FANP) Criticality Analysis report(Ref. 4) lists the NUHOMS
-32PT DSC Type A, B, and C basket designs and soluble boronrequirements for the FCS fuel. A Type A fuel basket with a 16-poison plate configuration is the most reactive fuel basket design because of the fewer poison plates. Although FCS spent fuel may be stored in the Type A, B, or C baskets, for conservatism, the Type A basket with 16 poison plates and no credit for PRAs is used in the criticality analyses performed in support of this amendment request. 3.2Description of Proposed Technical Specification ChangesAttachment 2 to the November 8, 2005, letter provided a markup of the TSs and correspondingbases pages. The following is the descriptive list of the changes proposed by the amendment request:(1) Addition of TS 2.8.3(6), "Spent Fuel Cask Loading" TS 2.8.3(1) through 2.8.3(5) in the existing TS LCO 2.8.3, "Refueling Operations - Spent FuelPool," include the limits and controls for storage of unirradiated fuel and spent fuel in the FCS spent fuel storage racks. The licensee proposed to add TS LCO 2.8.3(6), "Spent Fuel CaskLoading," specifying (1) a new minimum boron concentration limit of 800 parts per million (ppm)for the SFP during spent fuel cask loading operations, and (2) a new burnup versus enrichment curve (TS Figure 2-11) for fuel assemblies located in the spent fuel cask in the SFP.
TS 2.8.3(6) also includes required actions when these LCOs are not met.The acceptability of the new LCO 2.8.3(6) is demonstrated by the licensee's criticality analysisfor compliance with the subcriticality criteria of 10 CFR 50.68. The staff evaluation of the criticality analysis is described in Section 3.3 of this report.(2) Revision to TS Table 3-4, "Minimum Frequencies for Sampling Tests" The existing Footnote (4) for TS Table 3-4 specifies the sampling test frequency of the SFPboron concentration as "prior to placing unirradiated fuel assemblies in the spent fuel pool and weekly when unirradiated fuel assemblies are stored in the spent fuel pool." The licensee proposed to revise Footnote (4) to "prior to placing unirradiated fuel assemblies in the spent fuel pool or placing fuel assemblies in a spent fuel cask in the spent fuel pool, and weekly whenunirradiated fuel assemblies are stored in the spent fuel pool, or every 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> when fuelassemblies are in a spent fuel storage cask in the spent fuel pool." [Underline added] As a result of this amendment request, this revision is needed to address the boronconcentration sampling test frequency for the conditions when spent fuel assemblies are located in a spent fuel cask in the SFP. The proposed change provides a new restriction for the spent fuel cask that is more restrictive than the current TSs and, therefore, the change isacceptable.(3) Revision to TS Table 3-5, "Minimum Frequencies for Equipment Tests" The licensee proposed to add new item 24 to TS Table 3-5 to address spent fuel cask loadingoperations by requiring the licensee to "verify by administrative means that initial enrichment and burnup of the fuel assembly is in accordance with Figure 2-11" at a frequency of "prior to placing the fuel assembly in a spent fuel cask in the spent fuel pool."This revision is an added surveillance requirement to assure that LCO 2.8.3(6)(2) regarding TSFigure 2.11, burnup-enrichment load curve, is met. The proposed change provides a new restriction for the spent fuel cask that is more restrictive than the current TSs and, therefore, thechange is acceptable. (4) Revision to TS 4.3.1, "Criticality for Fuel Storage" The licensee proposed to add TS Subsection 4.3.1.3 to specify that the spent fuel casks aredesigned and shall be maintained with: a.Fuel assemblies having a maximum U-235 enrichment of 4.5 weight percent, b.keff < 1.0 if fully flooded with unborated water, which includes an allowance foruncertainties as described in Section 9.5 of the USAR [updated safety analysis report], c.keff 0.95 if fully flooded with borated water 800 ppm, which includes an allowance foruncertainties as described in Section 9.5 of the USAR, d.A nominal 9.075-inch center-to-center distance between fuel assemblies placed in thespent fuel cask, e.Spent fuel assemblies with a combination of discharge burnup and initial averageassembly enrichment in the "acceptable" range of Figure 2-11.This added Section 4.3.1.3 describes the characteristics of the spent fuel cask loading in theSFP. Items b and c specify the commitment to meet 10 CFR 50.68 reactivity limits. Items a, d, and e, respectively, specify the requirements regarding the enrichment limit, fuel loading spacing, and minimum burnup limit of fuel assemblies loaded in the spent fuel cask to assure compliance with the subcriticality requirements. The proposed change provides new TS restrictions for the spent fuel cask which previously did not exist and, therefore, the change ismore restrictive and is acceptable.(5) Editorial Changes As result of the proposed TS changes, several editorial changes to the formatting of the TSswere needed. In the Table of Contents, the Figure 2-11 was added and the other figures wereplaced in numerical order. Also on this page, the spelling of "concentration" in the title ofFigure 2-3 was corrected. In Section 2.8, Pages 14, 15, 26 and 27 were added. Because pages 14 and 15 were new, Section 2.8 from pages 14 to 23 were repaginated to 16 to 25.
Section 4.3.1.3 was added to Section 4.0 on page 2. As a result some text on page 2 was moved to page 3. The staff has reviewed these changes and concluded that they are editorial in nature and, therefore, are acceptable. In our review of the proposed TS changes, the NRC staff reviewed the supporting criticalityanalysis to ensure that it is consistent with the assumptions and inputs used in the supportingsafety analysis.3.3Criticality AnalysisAttachment 1, "Criticality Control During Spent Fuel Cask Loading in the Spent Fuel Pool," andEnclosure 1, "Framatome ANP [Framatome Advanced Nuclear Power] Criticality Analysis," to the November 8, 2005, letter, as amended by LIC-06-0023 Enclosure 1, "NUHOMS-32PT 50.68Criticality Analysis for Fort Calhoun (Revision Number 1)" (Ref. 4), describe the criticality analyses performed by the licensee and its contractor FANP to support the proposed amendment for the spent fuel cask loading in the NUHOMS-32PT DSC/TC in the FCS SFP. The FANP modeling approach used bounding parameters to produce the maximum keff. Indetermining the acceptability of the amendment, the NRC staff evaluated the appropriatenessof the computer codes employed for the analyses, the methodology used to calculate themaximum keff, and the criticality analyses to demonstrate compliance with the regulatoryreactivity limits. The NRC staff evaluated whether the licensee's analyses and methodologiesprovided reasonable assurance that adequate safety margins in accordance with NRC acceptance criteria were developed and could be maintained in the FCS SFPs during cask loading, unloading, and handling operations. 3.3.1Criticality Analysis Codes The FCS spent fuel cask loading criticality analyses were performed using the KENO V.a codeto calculate keff. The CASMO-3 code was used for the fuel depletion analyses anddetermination of the isotopic atom densities of the spent fuel assemblies. The FANP KENO V.acode is a part of the SCALE Version 4.4a code package (Ref. 5) operating on the Linux operating system platform. KENO V.a is a three-dimensional Monte Carlo criticality code thatwas benchmarked against critical experiments that are representative of SFP and canister configurations. CASMO-3 is a multi-group, two-dimensional transport theory program for burnup calculations on light-water reactor assemblies consisting of cylindrical fuel rods of varying composition on the square array. It is typically used to generate cross-sections of the fuel cycles, and is used for reactivity studies to provide depletion data for burnup credit. Thesecodes are widely used for the analysis of fuel rack reactivity and have been benchmarked against results from numerous critical experiments. In an NRC memorandum dated August 19,1998, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," (Ref. 3) providing guidance on fuel storage criticality analysis, the NRC staff stated that KENO V.a and CASMO-3 codes were acceptable computercodes for the analysis of fuel assemblies stored in the SFP. The licensee performed independent benchmarks to establish the appropriate uncertainties forthe criticality safety analysis of the NUHOMS-32PT DSC in the FCS SFP, includingcomparisons of the FANP KENO V.a calculations to those from TN calculations that were independently benchmarked to a set of critical experiments. Also, the KENO V.a calculations were used to model the FCS SFP in support of earlier TSs. The CASMO-3 calculations were used for benchmark comparisons of the FCS reactor operation during several reload cycles. Table 4.4 of Enclosure 1 provides a comparison of an FANP KENO calculation result of keff and for different neutronic histories ranging from 500,000 to 4 million. The table shows the resultsare statistically equal for all cases, except for a couple of cases that slightly exceed 2. Therefore, cases with approximately 1 million histories are sufficient to assure convergence ofthe KENO V.a reactivity calculations. However, FANP elected to use 2.5 million histories for conservatism. The NRC staff concludes that the analysis methods used are acceptable andcapable of predicting the reactivity of the FCS spent fuel casks in the SFP with a high degree ofconfidence.3.3.2Criticality Analysis Methodology The NRC Criticality analysis guidance (Ref. 3) specifies that the Maximum keff value for thecriticality analysis should be the summation of the KENO V.a-calculated keff, the bias in criticalityanalysis methods, manufacturing and calculational uncertainties, and the correction for the effect of the axial distribution in burnup when credit for burnup is taken. Uncertainties were determined for the proposed storage facilities and fuel assemblies to account for tolerances in the mechanical and material specifications. An acceptable method for determining themaximum reactivity may be either (1) a worst-case combination with mechanical and material conditions set to maximize keff, or (2) a sensitivity study of the reactivity effects of tolerancevariations. The licensee's modeling approach for the NUHOMS-32PT DSC/TC criticality analysis is theuse of bounding parameters in the KENO V.a design-basis model that produces the maximum
keff. The KENO model of the DSC/TC utilized in the criticality analysis is directly obtained fromthe bounding criticality analysis model utilized in the Part 72 criticality analysis. The design-basis model is constructed utilizing the worst case geometric and material tolerances that resultin the most conservative calculation of the system reactivity. A series of sensitivity studies wereperformed to evaluate the reactivity effects of geometry and material tolerances of various DSC/Cask components and fuel. The FANP criticality analysis report presented the results of some sensitivity studies on the KENO V.a calculation of the keff, such as the effect of replacingthe lead (Pb) gamma shield in the transfer cask with stainless steel 304 of the "light" TC (Table 4-5), and the effect of eccentric positioning of fuel assemblies in the DSC fuel basket (Table 4-6). The design-basis criticality model is then constructed by combining deterministically the worst reactivity effects of these parameters. For the NUHOMS
-32PT DSC, the design-basis criticality analysis is modeled with the mostreactive system configurations, including the moderator density and fuel assembly spacing. The KENO V.a results include the bounding uncertainties for the canister and SFP. Since the uncertainties associated with the geometrical/material tolerances are conservatively built into the criticality analysis model of the DSC/TC, the only additional uncertainties that need to be considered are those associated with the methods and those associated with the burned fuelisotopic concentrations. KENO V.a Benchmark BiasThe FANP criticality analysis of the spent fuel cask loading in the SFP includes two sets ofindependent benchmarks: (1) experiments of specific configurations that are comparable to the FCS SFP and the DSC; and (2) experiments of fuel assembly configurations that are comparable to the FCS fuel assemblies.The KENO V.a benchmark calculations modeled 100 critical experiments that arerepresentative of SFP and canister configurations. Twenty-one of the 100 experiments wereperformed by FANP-B&W, and were recommended in the NRC guidance for criticality analysis(Ref. 3). This set included the effects of burnup by modeling plutonium fuel in addition to uranium fuel. The approach utilized for the NUHOMS
-32PT DSC criticality safety analysis wasto bound all uncertainties associated with the benchmark results. The bounding benchmark uncertainty was determined to be 0.01549 keff for criticality safety evaluations of loading theDSC in the FCS SFP.The second set of benchmarks validated the CASMO-3 methods. The results of thisbenchmark indicated that the bias and random uncertainties associated with CASMO-3 were smaller than those associated with KENO, i.e., no statistical 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 uncertainty was assigned to the CASMO methods with fuel assemblies. Spent Fuel Burnup UncertaintiesIn the burnup credit analyses, biases are added to the bounding criticality safety predictions toaccount for the uncertainties associated with (1) the axial burnup distribution in the active fuel length, and (2) the assembly burnup accuracy. The dominant bias is from the axial burnup effects. The burnup credit calculations were performed assuming a uniform burnup profile throughoutthe active fuel length that results in the over-prediction of burnup at the ends of the fuelassembly and under-prediction of burnup in the fuel mid-region. The "axial end-effect" bias, i.e., the difference of the keff values between the axial burnup profile and the uniform burnupassumption, needs to be applied to the burnup credit calculations to account for the increase in reactivity. Generic analyses confirmed minor and generally negative reactivity effect of the axially distributed burnup at values less than 30,000 megawatt days/metric tons Uranium (MWD/MTU). As a result, KENO calculations with less than 30,000 MWD/MTU do not contain an axial bias. The highest burnup evaluated in the analysis was 38,000 MWD/MTU, to which the axial bias uncertainty of +0.013 k is applied.Another major contributor to the bounding uncertainty is the bias in the assembly burnup toaccount for the uncertainty in the fuel depletion calculations. The NRC guidance document(Ref. 3) stated an uncertainty equal to 5 percent of the reactivity decrement to the burnup of interest is an acceptable assumption. The licensee in its response to an NRC staff question(Ref. 2) stated that the reactivity decrement uncertainty due to burnup was incorporated by shifting the burnup curve upwards by 5 percent uniformly for all burnups.In summary, the NRC staff evaluated the licensee's criticality analysis methodology, andconcluded that the use of bounding parameters to produce the maximum keff is consistent withthe NRC criticality analysis guidance and is acceptable.
3.3.3DSC Loading Criticality Analysis To demonstrate compliance with the reactivity criteria of 10 CFR 50.68, the licensee performedthe criticality analyses of the NUHOMS-32PT DSC/TC cask loading in the SFP for both normaland accident conditions. The criticality analyses were performed with the NUHOMS-32PTDSC Type A fuel basket with 16 poison plates and no PRA, which is the most reactive and bounding DSC basket design. Consistent with the methodology of using bounding parameters,the criticality analyses were performed with the conservative assumptions that tend to maximize the reactivity. These assumptions include: *FCS SFP peripheral cells adjacent to Cask Pit Area maintained empty during DSCloading. *No burnable poisons accounted for in any fuel assembly in the KENO model.
- The transition rails between the basket and the canister shell modeled as 100 percentaluminum (steel in the transition rails reduces reactivity because of much higher absorption cross-sections). *Control rods assumed inserted for CASMO depletion calculations to maximize keff. *Water density was at the optimum moderator density of 1.00 gram/cc, corresponding to 4 o C. *All fuel rods are filled with fresh water in the pellet/cladding gap.
- Full DSC reflection in the radial direction.
- Only 90 percent of credit taken for of B-10 in the DSC neutron poison plates.
OPPD has demonstrated compliance with the subcriticality criteria of 10 CFR 50.68 that keff 0.95 if credit is taken for soluble boron in the SFP, and keff < 1.0 if flooded with fresh water. Two normal spent fuel cask loading operation conditions were analyzed with fresh water and 500 ppm borated water, respectively, in the SFP. For the case of the SFP flooded with fresh water, the licensee performed a series of CASMO/KENO runs with various enrichment-initial burnup combinations to achieve the maximum keff < 1.0. The results, shown in Table 6-1 andFigure 6-1 of the FANP criticality analysis report, show that the most reactive configuration has
keff = 0.99713, below the acceptance criterion of 1.0. It should be noted that the assemblyburnups listed in Table 6-1 and Figure 6-1 have been increased by 5 percent from the burnups used in the analysis to account for burnup uncertainty. This analysis shows that a complete dilution of the boron in the SFP would not cause a criticality in the NUHOMS
-32PT DSC. Themost reactive case was again analyzed with 500 ppm soluble boron in the SFP. The results show keff = 0.92073, below the acceptance criterion of 0.95.The licensee analyzed two criticality accident conditions, i.e., misloading of a fresh fuelassembly and dropping of a fresh fuel assembly, that are consistent with the FCS current design and licensing basis for the SFP. The accident analyses were analyzed based on the "double contingency" principle that at least two unlikely, independent accidents or upsets haveto occur concurrently or sequentially for a criticality event to be possible. Therefore, a boron dilution event in the SFP, that is not a design-basis event for the FCS, was not postulated tooccur as an accident event, either individually or concurrently with the two accidents analyzed.
The fuel misloading accident was analyzed assuming that the transfer cask was loaded with fuel assemblies having enrichments up to 4.55 wt% U-235, and the empty position loaded with a fresh fuel assembly of 4.5 wt% 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 were assumed in the analysis. The results, provided in Table 6.3 of the criticality analysis report, show that the minimum required soluble boron concentration is 800 ppm to maintain keff 0.95 with all uncertainties. The bounding value for the misloading fuel bundleaccident with 800 ppm soluble boron is 0.94623 for the DSC fuel enrichment of 4.55 wt%.The fuel drop accident was evaluated assuming the dropped assembly was placed along theperimeter of the transfer cask aligned longitudinally and evaluated at different azimuthallocations to find the most reactive position. The analysis was performed for various soluble boron concentrations. The results, shown in Table 6-4 of the criticality analysis report, show that with the 800 ppm boron concentration, the most reactive case has keff = 0.92499. The results of the criticality analyses for normal and accident conditions are summarized inTable 4.1-1 of Attachment 1 to the November 8, 2005, letter. The results show that the spentfuel cask system remains sufficiently subcritical under normal and applicable accidentconditions in the FCS licensing basis. Based on the results of the criticality analyses, a minimum boron concentration of 800 ppm is required in the DSC fuel cavity water during fuel loading for accident conditions. In addition, Table 4.1-2 of Attachment 1 adopts from Table 6.1of the criticality analysis report the burnup-enrichment requirement. All FCS fuel assemblies loaded into the NUHOMS-32PT DSC must have a minimum average assembly burnup greaterthan or equal to the value shown in Table 4.1-2.Since the criticality analyses for normal and accident conditions of the DSC loading with fuelassemblies, with the burnup-enrichment combinations of Table 4.1-2, have shown compliance with the subcritical criteria of 10 CFR 50.68, these burnup/enrichment combinations are used asthe burnup credit loading curve. This loading curve is adopted as TS Figure 2-11, "Limiting Burnup Criteria for Acceptable Storage in Spent Fuel Pool."
The NRC staff reviewed the licensee's amendment to add LCO 2.8.3(6) specifying (1) aminimum boron concentration requirement of 800 ppm for the SFP during spent fuel cask loading operations, and (2) a new burnup versus enrichment curve for fuel assemblies located in the spent fuel cask, as well as other changes to TS Tables 3-4 and 3-5, and TS 4.3.1, relatedto the spent fuel cask loading operations in the FCS SFP. In addition, editorial changes were made mostly to make the TSs consistent with the proposed changes and to conform the pagination. Based on its review of the licensee's criticality analysis, the NRC staff concl udesthat the proposed changes for the spent fuel cask loading operations in the SFP meetappropriate subcriticality requirements of 10 CFR 50.68 and GDC 62. The NRC staff f ound thatthe licensee's amendment request provided reasonable assurance that under both normal and accident/upset conditions, that licensee would be able to safely operate the plant and complywith the NRC regulations. Therefore, the NRC staff finds the amendment acceptable.
4.0STATE CONSULTATION
In accordance with the Commission's regulations, the Nebraska State official was notified of theproposed issuance of the amendment. The State official had no comments.
5.0ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facilitycomponent located within the restricted area as defined in 10 CFR Part 20. The NRC staff hasdetermined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there isno significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (70 FR 75494; December 20, 2005). Accordingly, the amendment meets the eligibility criteriafor categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) thereis reasonable assurance that the health and safety of the public will not be endangered byoperation in the proposed manner, (2) such activities will be conducted in compliance with theCommission's regulations, and (3) the issuance of the amendment will not be inimical to thecommon defense and security or to the health and safety of the public.
7.0REFERENCES
1.Letter from R.T. Ridenoure (OPPD) to the U.S. Nuclear Regulatory Commission, "FortCalhoun Station Unit No. 1 License Amendment Request (LAR)05-013, 'Criticality Control During Spent Fuel Cask Loading in the Spent Fuel Pool,'" November 8, 2005, LIC-05-0119.2.Letter from D. J. Bannister (OPPD) to the U.S. Nuclear Regulatory Commission,"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,'" March 27, 2006, LIC-06-0023.3.NRC Memorandum from L. Kopp to T. Collins, "Guidance on the RegulatoryRequirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," August 19, 1998. 4.LIC-06-0023 [Reference 2 above] Enclosure 1, "NUHOMS-32PT 50.68 CriticalityAnalysis for Fort Calhoun (Revision Number 1)."5.NUREG/CR-0200, "SCALE: A Modular Code System for Performing StandardizedComputer Analyses for Licensing Evaluation," distributed by the Radiation ShieldingInformation Center, Oak Ridge National Laboratory, Oak Ridge, Tennessee, September
1998.Principal Contributor: Y. Hsii Date: April 10, 2006 January 2006Ft. Calhoun Station, Unit 1 cc:
Winston & Strawn ATTN: James R. Curtiss, Esq.
1400 L Street, N.W.
Washington, DC 20005-3502ChairmanWashington County Board of Supervisors
P.O. Box 466 Blair, NE 68008Mr. John Hanna, Resident InspectorU.S. Nuclear Regulatory Commission
P.O. Box 310 Fort Calhoun, NE 68023Regional Administrator, Region IVU.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005Ms. Julia Schmitt, Manager Radiation Control Program Nebraska Health & Human Services R & L Public Health Assurance 301 Centennial Mall, South P.O. Box 95007 Lincoln, NE 68509-5007Mr. David J. Bannister, ManagerFort Calhoun Station Omaha Public Power District Fort Calhoun Station FC-1-1 Plant
P.O. Box 550 Fort Calhoun, NE 68023-0550Mr. Joe L. McManisManager - Nuclear Licensing Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.
P.O. Box 550 Fort Calhoun, NE 68023-0550Mr. Daniel K. McGheeBureau of Radiological HealthIowa Department of Public HealthLucas State Office Building, 5th Floor 321 East 12th Street Des Moines, IA 50319