Revision as of 15:57, 5 April 2018

Text

Indian Point, Units 2 and 3 - Request for Additional Information Regarding Spent Fuel Transfer
	ML11243A175
Person / Time
Site:	Indian Point
Issue date:	08/23/2011
From:	; Entergy Nuclear Operations
To:	; Office of Nuclear Reactor Regulation
References
	NL-11-100, TAC L24299, TAC ME1671, TAC ME1672
	Download: ML11243A175 (94)
	v • d • e

HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONREQUEST FOR ADDITIONAL INFORMATIONREGARDING SPENT FUEL TRANSFERENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3DOCKET NOS. 50-247 AND 50-286By letter dated July 8, 2009 (Agencywide Documents Access and Management System(ADAMS) Accession Nos. ML091940177 and ML091940178), and supplemented by lettersdated September 28, 2009, (ADAMS Accession Nos. ML092950437 and ML093020080), andOctober 5, 2010 (ADAMS Accession Nos. ML1 02910511, ML1 03080112, and ML1 03080113)Entergy Nuclear Operations, Inc. (Entergy or the licensee), submitted a license amendmentrequest for Indian Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3). The proposedchanges are requested to provide the necessary controls and permission required for Entergy tomove spent fuel from the IP3 spent fuel pool (SFP) to the IP2 SFP using a newly designedshielded transfer canister (STC), which is placed inside a HI-TRAC 1OOD cask for outdoortransport. The chapters listed below refer to the safety analysis report (SAR) for the STC,HI-2094289, Revision 3, ADAMS Accession No. ML103080113. The Nuclear RegulatoryCommission (NRC) staff is reviewing the submittal and has the following questions.The following provides the NRC questions together with Entergy's responses.CHAPTER 1 -GENERAL INFORMATIONNRC RAI 1-1Attachments 5 and 7 to your letter dated October 5, 2010, included a proposed Appendix C tothe unit operating licenses, with technical specifications (TSs) for the spent fuel transferoperations. Proposed TS 3.1.3, "Shielded Transfer Canister (STC) Pressure Rise," included alimiting condition for operation (LCO) of less than a 4.2 psi pressure increase over 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .The associated proposed TS Bases were provided in Attachments 6 and 8, and the basesindicated that the LCO ensures that fuel assemblies selected for loading in the STC satisfydesign basis limits because an analysis of the pressure rise for the design heat load of 9.6 kWwas less than 4.2 psi. However, the licensing report (Enclosure 1 to letter dated October 5,2010) includes additional analyses verifying that the STC design pressure would not beexceeded for various postulated accident conditions. (SBPB)Paragraph (c) (2) of Title 10 of the Code of Federal Regulations (10CFR) 50.36, "TechnicalSpecifications," requires in part that a LCO be established for a process variable, designfeature, or operating restriction that is an initial condition of a design-basis accident or transientanalysis that either assumes the failure of or presents a challenge to the integrity of a fissionproduct barrier. In addition, Paragraph (c)(3) of 10 CFR 50.36 requires the inclusion ofSurveillance Requirements (SRs) related to test or inspection that assure that the LCOs will bemet. Finally, Paragraph (c)(4) of 10 CFR 50.36 requires the inclusion in the TSs of designfeatures of the facility that, if altered or modified, would have a significant effect on safety.The NRC staff considers the STC pressure boundary, which is identified as a confinementboundary, to be a fission product boundary. The licensing design-basis analyses examine thePage 1 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONeffects of several postulated events with respect to challenges to the integrity of the fuelcladding and the STC pressure boundary. Accordingly, propose additional LCOs, SRs, andDesign Feature TSs necessary to satisfy the requirements of 10 CFR 50.36.For example, address methods to verify that appropriate initial conditions have been establishedfor STC overpressure protection (e.g., a steam bubble of appropriate volume exists within theSTC). To ensure this steam space is formed, additional controls are needed. Proposed TS4.1.4.6 should be converted to a TS SR and LCO. This SR would verify that the requiredvolume of water is removed through the STC drain line following the application of steampressure. Also, in the absence of an analysis showing the acceptability of an air-filled spacerather than a steam-filled space, a TS SR needs to demonstrate that the space is filled withsteam. One method would be to use the pressure change in the STC during the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months sfollowing establishment of the steam space. LCO 3.1.3 could be modified to show that thepressure stays within an analyzed range over time (e.g., a response graph). The lower limit ofthe graph could be the zero heat load curve, and the upper limit could be the response with thedesign heat load. See RAI TS-8 for additional comments on LCO 3.1.3.Design features essential for the assumed heat transfer capabilities necessary to preventoverpressure conditions (e.g., materials of construction [for thermal conductivity] and emissivityof outer surface) need to be described in TS. For example, TS section 4.1 could be modified toadd a section (after Criticality) titled "4.1.3 Thermal Features." The type of thermal featuresdescribed here would be those critical to the heat transfer abilities of the STC and the HI-TRAC.This information is needed to confirm compliance with 10 CFR 50.36.Response to RAI 1-1The proposed LCOs and SRs have been revised to satisfy the requirements of 10 CFR 50.36.The STC pressure boundary is a confinement boundary and therefore, by definition, a fissionproduct boundary. The accident analyses assume that the STC initial water level has beenestablished, that the STC void space is filled with steam, and that the STC contains a designbasis heat load. Therefore, LCOs are proposed that establish these design basis conditionsprior to transfer operations.STC Initial Water LevelThe originally proposed TS 4.1.4.6 has been converted to TS LCO 3.1.3 STC Initial Water Leveland SR 3.1.3.1. The SR verifies that steam is emitted through the STC drain line and that therequired volume of water is removed following the application of steam pressure.Demonstration that the STC void space is filled with steam and detection of a severe fuelmisloadAs suggested in the RAI one method to demonstrate that the STC void space is filled withsteam would be to use the pressure change in the STC during the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months followingestablishment of the steam space wherein an LCO and SR could be developed to show that thepressure stays within an analyzed range over time. The lower limit of the graph could be thezero heat load curve, and the upper limit could be the response with the design heat load. RAITS-8 also suggests that 25 data points be taken over a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period to allow more accuratechecking of the pressure change.Page 2 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONTo use upper and lower bound pressure or pressure rise measurements on an hourly basis asan LCO requirement for fuel transfer would be problematic for a number of reasons. Thepredicted thermodynamic response of the loaded STC is a complex issue that is a function ofthe accuracy of the thermodynamic model, a number of variables, and boundary conditions.These include, but are not limited to, pool water temperature, ambient air temperature, length oftime required to load fuel in the STC, length of time required to transfer the loaded STC from theSpent Fuel Pool to the HI-TRAC. Each one of these variables would need to be measured andrecorded in order to determine the proper pressure or pressure rise curve to use to confirm thatthe heat load of the fuel is below the design basis limit and that the STC void space is filled withsteam.Table 1 below shows how the pressure rise inside the STC will change as a function of avariation in the initial temperature in the STC. There are multiple other variables that also canaffect the STC pressure or pressure rise, in both a positive and negative direction.Table 1: 24 Hour Pressure Rise for Desigqn Basis Heat Load, 100°F Ambient TemperatureGiven the uncertainties in the thermal model predictions and pressure readings, it is notpractical to develop an LCO and associated SR that would require the comparison of predictedand measured pressure or pressure rise on an hourly basis. Such a comparison could lead toboth unwarranted fuel handling and transfer delays.Based on the above discussion it is also not practical to attempt to distinguish between asignificant misload and significant amounts of air in the STC steam space as both conditionsresult in an STC pressure rise greater than that predicted by the thermal analysis for designbasis conditions. Therefore, an LCO and associated SR are proposed that would detect eitherof these conditions and require specific actions to be taken in the event that the rate of STCpressure rise, over a rolling four hour period, exceeds the specified rate of change. A rolling 4hour period, with measurements taken hourly for a total of 25 data points over a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period,was selected to ensure that uncertainties in the thermal analysis and pressure readings do notresult in unwarranted actions and also to ensure that there is sufficient time available to mitigateeither condition (significant fuel misload or significant amount of air).As documented in Chapter 5 of the Licensing Report, the 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months STC pressure rise has beendetermined based on a design basis heat load and with steam in the STC void space (Figure 1below). The STC pressure rise for a severe misload is compared to the design basis pressurerise in Figure 2 and likewise the rate of change of STC pressure are compared in Figure 3. ThisFigure shows that a rate of change of be used todifferentiate between a design basis heat load and a severe misload shortly aftercommencement of the STC pressure rise surveillance. This criterion also serves to detect asignificant amount of air inside the STC, since that condition would also substantially increasethe rate of change of STC pressure.A rate of STC pressure rise above rpR ARY TEXT REMOVE Dduring any rolling 4hour period is an indication that the design basis of the STC is not being met. This indicationcould represent one or more of the following conditions: not establishing the correct STC waterlevel, misloading a fuel assembly, or a significant ingress of air into the STC. In this case theSTC would be depressurized and actions taken to verify: the STC water level, that a fuelPage 3 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONmisload had not occurred and that air had been effectively precluded from entering the STC.Once the cause of the non conforming condition is identified and corrective actions taken, theSR would need to be re-performed satisfactorily prior to STC transfer operations.In support of the proposed LCO and SR it should be noted that the STC is designed forpressures well above those required to demonstrate compliance with this LCO. Calculations, asdocumented in Chapter 5 of the Licensing Report, have been performed to demonstrate that theSTC pressure boundary can accommodate pressures of up to 90 psig, which isk RPX10V F.ED] more than the steady state pressure for design basisheat loads, and two times more than the steady state pressure for [PROPRIEtmýY -ýftT,EMOVEbD. Therefore, although the LCO and SIR will preclude it, the STC could be loadedwith heat loads even greater [ý.P itARY]before the steady statepressure would exceed the pressure limits for the STC. In addition, once compliance with thisLCO is established, it is judged that there are no other credible design basis accidents duringthe transfer which challenge the pressure boundary of STC. The design basis accidents aredescribed in Chapter 5 of the licensing report and include a loss of jacket water, a fire, HI-TRACannulus plus a jacket water loss, fuel misload, and a tip-over accident. Apart from the tip-overaccident there are sufficient margins to include any minor misload not detectable via theproposed LCO and surveillance. While the tip-over accident has been analyzed, there is nocredible mechanism for this to take place and it therefore considered to be a non-mechanisticevent and demonstration of a misload coincident with a tipover is not considered credible.tkR" 'tOE Y T MOVE.The STC pressure rise determined based on full STCdesign basis heat, pool water temperature of 120°F and a significant amount of air entrapmentin the STC 'TEXT REMOVED]. Therefore, the [POPZI FTARY TEkiIRENTOVEDjcriterion over a rolling 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period ensures that a severe fuel misload orsignificant amounts of air in the STC would be detected by the pressure monitoring systemwithin 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .[PROP RIE 'TARY TEIFXT REFMOVEDIFigure 1: 48-hour STC Pressure Rise under Design Basis Heat Load[PROPRIETARY 'TEXT REM.(DY ED]Figure 2: Comparison of STC Pressure Rise between Design Basis Heat Load and Severe FuelMisload Accident(PROPRIETARY TEXT REM.OVEDI"Figure 3: Comparison of Rate of Change of STC Pressure with Time between Design BasisHeat Load and Severe Fuel Misload AccidentThe RAI also requests that "thermal features" be added to the TS. The thermal model is basedon the nominal dimensions provided in the licensing report and explicitly considers all thefeatures (such as gaps between basket and the STC shell) that affect the thermal-hydraulicbehavior of the STC system. Importance of individual design features of the STC in determiningthe thermal-hydraulic behavior of the STC cannot be quantified and measured. In order toconservatively predict the temperature/pressure in the STC and HI-TRAC, an array ofconservatism assumptions have been used in the analyses (see section 5.3.1 of the LAR).Page 4 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONConservative material properties from valid references have been used in the analyses. Nospecial treatment of the materials is required to ensure their thermal capabilities or meet thevalues assumed. Also note that the following statement has been added to the TS Bases B3.1.4, "...24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months STC pressure test will also be a check on the thermal properties of the STCand HI-TRAC system, such that if there were any significant changes in the thermal properties itwould be expected to show as anomalous readings during the pressure test." Therefore, furthercontrols in the TS on the thermal features are not needed.NRC RAI 1-2Modify licensing drawing No. 6013, Sheet 2 of 4, to state that the lead thickness shown is aminimum thickness. (CSDAB)The lead thickness for the STC in the referenced drawing is the minimum thickness allowed perthe proposed TS, Appendix C, Part I, Section 1.0 description of the STC; thus, the drawing'sspecification of the lead thickness should be consistent with that description and indicate thatthe stated dimension in the drawing is a minimum value.This information is needed to confirm compliance with 10 CFR Part 50, Appendix B, and theintent of 10 CFR 72.104, 72.106 and 72.44(a).Response to RAI 1-2Drawing 6013 (Sheet 2 of 4) has been revised to state that the lead thickness shown is aminimum thickness (Section 1.5 in the Licensing Report).CHAPTER 4 -CRITICALITY EVALUATION (CSDAB and SRXB)General Response to Criticality RAIsBefore the detailed responses to the individual RAIs are presented, this introductory sectiongives an overview of the criticality methodologies, regulatory bases, applicable acceptancecriteria, embedded margins and conservatisms, and discusses general adjustments that weremade as a result of the RAI responses.Governing Regulation, B-10 Areal Density, Initial MethodologyThe STC is a device that is to be licensed under 1OCFR50, and its criticality safety performanceand acceptance criteria are therefore governed by 1 OCFR50.68. The design of the STC basketis based on the specifications of the basket of the MPC-32 dry storage and transportationcanister. The STC basket consists essentially of the 12 inner cells of an MPC-32 basket. Froma criticality perspective, the STC has a larger neutron absorption capability than almost all otherwet storage structures (racks) licensed under 1OCFR50.68. Specifically, the B-10 areal densityand steel wall thicknesses are substantially larger than those in the IP Unit 3 and Unit 2 Region2 spent fuel racks (See Licensing Report Section 2.2.1 and Table 2.2.1). Based on this factPage 5 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONalone, all fuel qualified for the Region 2 racks in the IP Unit 3 and Unit 2 pool wouldautomatically qualify for loading into the STC. Specifically," The B-1 0 areal density of the neutron absorbers in the STC basket is 0.0310 g/cm2(minimum), which is higher than that of typical wet storage racks in the US. The IP2 andIP3 racks have neutron absorbers with nominal B-10 areal densities of 0.026 and 0.020g/cm2, respectively.* The thickness of the wall of the steel basket is about 0.28 inch, while typical wet storageracks have thicknesses of about 0.075 inch. The wall thickness in the IP2 and IP3 racksare 0.075 and 0.170 inches, respectively.Based on this situation, the initial licensing report for the STC utilized a typical wet storagemethodology to demonstrate criticality safety, with soluble boron credit for accident conditionsbut not for normal conditions, and applied essentially the same burnup requirements (loadingcurves) that existed for the IP2 and IP3 Region 2 racks to the STC.Change to HI-STAR 100 Criticality Methodology, Retaining 10CFR50.68 Acceptance CriteriaWhen responding to the first round of RAIs in 2010 for criticality, an additional aspect was takeninto consideration. Over the last two years, the general criticality analysis methodology for wetstorage systems has been undergoing an extensive review and revision. As of this writing, thisprocess is not finalized, and new durable NRC guidance is only expected in early 2013. In theinterim, a draft ISG (DSS-ISG-2010-01) was issued by the NRC in September 2010; howeverthat also was not available when the previous RAIs were received in April 2010. Therefore, inthe absence of a durable guidance, a methodology different from and more conservative thanthe typical Part 50 wet storage criticality methodology was used. This methodology is based onthe burnup credit methodology developed for the HI-STAR 100 Transport cask, which wasreviewed and approved by the NRC under 10CFR71 (ML062860201, October 12, 2006). TheHI-STAR criticality methodology introduced additional margin and conservatisms, mainly in thefollowing areas:* In addition to the traditional critical experiment benchmarks used in wet storage, the HI-STAR 100 methodology uses benchmarks based on chemical assays and commercialreactor criticals. Application of those additional benchmarks limits the number ofisotopes credited, applies highly conservative correction factors to minor actinides andfission products, and adds additional bias and bias uncertainties. While the Part 50 wetstorage criticality methodology combines uncertainties statistically, including thedepletion uncertainty, worst case combinations of tolerances are used in the HI-STAR100 methodology.* In the HI-STAR 100 methodology, assemblies with potential control rod insertion areconservatively analyzed as if control rods had been inserted to the fullest extent in thoseassemblies, although this condition is not permitted during full power operation. Thisapproach maximizes the spectrum hardening in those assemblies and thereforeincreases reactivity.It is important to note that using the Part 71 methodology vs. the Part 50 methodology is only achange in methodology, not the applicable regulations and acceptance criteria originally stated.The STC will still be licensed under 1OCFR50, and the acceptance criteria from 1OCFR50.68Page 6 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONstill apply. This means that soluble boron credit is an option in limiting k-eff under normalconditions. 10CFR50.68 also permits credit for soluble boron to protect against theconsequences of accidents, and this is applied in the analysis for the STC with respect tomisloading conditions. A burnup measurement to protect against misloading, as suggested inISG-8 Revision 2 as additional guidance to NUREG-1617 for applying burnup credit to transportpackages licensed under1 0 CFR Part 71, is therefore not necessary for the STC. A moredetailed comparison of the methodologies is presented below.Application of the HI-STAR 100 Methodology to the STCMost aspects of the burnup credit methodology for the STC are identical to those previouslydeveloped for the HI-STAR 100. However, there are some differences, either to account for site-specific conditions or in consideration of the different acceptance criteria (Part 71 vs. Part 50).This section provides a brief description of those differences, and provides or references therelevant discussions and justifications in each case.The differences are presented in the following tables, which briefly describe the aspects, anychanges made in the context of the current RAI responses, together with any discussion,justification, and references. The first table lists those aspects that were initially different (in-response to the first round of RAIs), and how the STC approach is now aligned with the HI-STAR 100 approach in the context of the current RAI responses. The second table presentsthose aspects where a difference remains, either in the initial calculations, or as a result of theRAI responses.Table 4.0.1- Resolved Methodology Differences HI-STAR 100 vs. STCHI-STAR 100 STC, First Submittal STC, Revised Discussion /using HI-STAR 100 Methodology for Justification /Methodology (2010) current RAI responses ReferenceBurnable Poison Rods Burnable Poison Rods Now same as HI-STAR Revised approach isassumed over entire over part of active 100 -Burnable Poison more conservative. Seeactive length length based on actual Rods assumed over Response to RAI 4-6burnable poison design entire active lengthused at IPBurnup Credit limited to Burnup credit exceeds Now same as rn-STAR Revised approach is50 GWd/mtU 50 GWd/mtU 100 -Burnup Credit more conservative. Seelimited to 50 GWd/mtU Response to RAI 4-8.Note that this is a limitof the credited burnup,not a limit of the actualburnup of an assembly.Table 4.0.2 -Retained and New Methodology Differences HI-STAR 100 vs. STCPage 7 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONHI-STAR 100 STC, First Submittal STC, Revised Discussion /using HI-STAR 100 Methodology for Justification /Methodology (2010) current RAI responses ReferenceUpper bound specific Same as HI-STAR 100 Conservatively low Revised approach ispower used in depletion specific power used in more conservative. Seeanalyses depletion calculations Response to RAI 4-5Radial Burnup Profiles Same as HI-STAR 100 Radial Burnup Profiles Revised approach isevaluated, but not now considered in more conservative. Seeconsidered in design design basis calculations Response to RAI 4-7basis calculationsNo optional loading Optional partial loading Unchanged Needed for fuel thatpattern for fresh fuel or pattern for fuel that does does not meet thefuel that does not meet not meet the loading minimum requiredthe loading curves. curves. burnup. Note that freshfuel is in fact notpermitted in the poolduring STC loading.Normal conditions Normal conditions Unchanged Note that IOCFR50analyzed with pure analyzed with pure regulations allows some(unborated) water (unborated) water soluble boron creditflooding, flooding, under normalconditions.Burnup Measurement to Soluble Boron Credit to Unchanged Standard approach forprotect against protect against I0CFR50 based onmisloading consequences of double contingencymisloading accidents. principle. Soluble boronprovides largesubcriticality margineven under misloadingconditions.Criticality Benchmarks Criticality Benchmarks Unchanged Increased confidence inbased on Fresh Fuel and based on Fresh Fuel, criticality calculationsMOX critical MOX fuel andexperiments simulated spent fuel(HTC) criticalexperimentNo additional Additional 5% B-10 Unchanged In consideration of theconsiderations for penalty for surveillance multipleneutron absorber program loading/unloadingsurveillance program cycles of the STCFor assemblies with As a bounding Unchanged Bounding approach topotential control rod approach, only insertion simplify analyses.Page 8 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONHI-STAR 100 STC, First Submittal STC, Revised Discussion /using HI-STAR 100 Methodology for Justification /Methodology (2010) current RAI responses Referenceinsertion, several during the entiredurations of this irradiation period isinsertions are consideredconsideredIn summary, for the differences presented in Table 4.0.2, the approach taken for the STCprovides the same or larger conservatism than that of the HI-STAR 100.Recent Part 50 Wet Storaqe Criticality ApprovalIt is also considered beneficial here to highlight recent developments in the wet storagecriticality safety analysis area: While there is no durable guidance yet, there are clear indicationsthat successful and acceptable paths forward for criticality safety evaluations for wet storagesystems are now available. Just recently, NRC approved the analysis for re-racking of theBeaver Valley Unit 2 Spent fuel pool, after a two year licensing process that involved extensiveinteraction with the NRC technical staff, including a week-long technical audit, and review ofanalyses by Oak Ridge National Laboratory experts (For the Beaver Valley SER seeML1 10890844). Beaver Valley is a Westinghouse plant with a similar fuel type and similarreactor conditions as Indian Point. It is noted that the burnup acceptance criteria developed andapproved for Beaver Valley is much less restrictive than developed for the STC basket using theHI-STAR 100 methodology. The STC will therefore provide a higher criticality margin than therecently approved Beaver Valley racks.Comparison of minimum required burnupsTo indicate the additional level of conservatism in the HI-STAR 100 methodology, the tablebelow lists the minimum required burnup for the fuel to be placed in the STC, in comparisonwith the minimum required burnup for fuel to be placed in the Indian Unit 2 and Unit 3 pools, andfor the Beaver Valley Unit 2 pool. For this comparison, all fuel is 4.0% enriched. As discussedabove, the Beaver Valley information is from a wet storage criticality license amendmentrequest approved very recently and is for fuel that is similar to the Indian Point fuel.Table 4.0.3 -Comparison of Minimum Required BurnupCalculation Minimum Required Burnup forfuel of 4 wt%- U-235 EnrichmentIGWd/mtU]Indian Point 3 Fuel in STCWet Storage Methodology (Initial 28.26Submittal)HI-STAR 100 Methodology, Fuel NOT 33.1 / 35.1exposed to control rods (First SubmittalPage 9 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONwith HI-STAR Methodology / Revisedfor RAI Responses)HI-STAR 100 Methodology, Fuel 40.5 /42.5POTENTIALLY exposed to controlrods (First Submittal with HI-STARMethodology / Revised for RAIResponses)Spent Fuel PoolsIndian Point Unit 2 28.80Indian Point Unit 3 29.75Beaver Valley Unit 2 (ML 110890844, 28.84Table 3.7.14-1E, Region 3)The minimum required burnup for the STC using the HI-STAR methodology is higher than eventhe requirements for the recently approved Beaver Valley pool. Again, this additionalconservatism to use the Part 71 HI-STAR 100 methodology was added intentionally to decouplethe STC criticality analyses from the currently developing wet storage methodology.Margin AnalysisThe HI-STAR 100 methodology is based on ISG 8, Rev 2. This ISG suggests performing amargin analysis to allow for an evaluation of any remaining uncertainties (or modelingdeficiencies) that are not explicitly considered in the design basis cases and compare againstthis margin. This margin analysis was performed for the HI-STAR 100, and also for the STC.For the STC, it indicates a potential margin on the order of 0.07 delta-k. With the revised loadingrequirements, there are essentially no uncertainties not explicitly considered in the analysis. Theentire margin is therefore available to cover any unspecified uncertainties.Indian Point Unit 3 Pool InventoryThe minimum required burnups are also viewed in the context of the actual inventory in theIndian Point Unit 3 pool that needs to be moved into the Unit 2 pool. The current requirements,with the substantial increase in minimum required burnups due to adopting the HI-STAR 100methodology (see Table 4.0.3), are fairly limiting in terms of the 12 assembly loadingconfiguration. Any further increase in the minimum required burnups would move a largernumber of assemblies into an 8 assembly loading configuration, thereby increasing the numberof transfers impacting both operations and ALARA considerations.Soluble Boron CreditThere is one difference between the actual condition of the STC and the condition that isassumed for the HI-STAR 100 for the criticality analyses: The STC is filled with borated waterwith a minimum soluble boron concentration of 2000 ppm, as specified in the proposedTechnical Specifications, whereas the HI-STAR 100 was analyzed with fresh water, as requiredby the regulations applicable to the HI-STAR 100 (Part 71).Page 10 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONThis soluble boron concentration is equivalent to an additional reactivity margin of about 0.20delta-k. The assumption of unborated water in the STC results in an additional and substantialmargin to any limit, i.e. a k-eff of 0.9 in unborated water would be a k-eff of about 0.7 with thisconcentration of soluble boron.The soluble boron level of the STC is controlled operationally through a proposed LCO in theTS, and since the system is designed and tested to be leak tight, there is no crediblemechanism to reduce this soluble boron level during transfer operations.Note that soluble boron, about 1000 ppm, is credited for various potential misloading conditions,as permitted by the regulation that governs the operation of the STC, 10CFR50.68. Thisregulation also permits credit for soluble boron under normal conditions, as long as themaximum k-eff remains below 1.0 for flooding with unborated water. This option is currently notapplied to the STC, and unborated water was used for the calculation to show that themaximum k-eff is below 0.95.Revised Loading CurvesRAI responses 4-5 through 4-7 required changes to the design basis calculations that resultedin revised loading requirements. Specifically considered were:* A conservatively low specific power is used in the depletion calculations instead of an upperbound power (see response to RAI 4-5)" Burnable Poison is now assumed over the full active length of the active fuel region, insteadof only over the reduced length of the burnable poison design (see response to RAI 4-6)" A conservative radial burnup gradient is now considered in all assemblies (see response toRAI 4-7)" All design basis calculations are performed with increased number of cycles (see responsesto RAI 4-2).* The maximum credited burnup is now limited to 50 GWd/mtU. This necessitates anenrichment limit for Configuration 1-B (see response to RAI 4-8)The minimum required burnup, and the calculated maximum k-eff values are listed in thefollowing tables.Parameter Configuration I AEnrichment, wt% U-235 2.0 3.0 4.0 5.0Bumup credited in 5.1 21.0 35.1 46.4Calculation, GWd/mtUMinimum required Burnup 5.4 22.1 36.9 48.7including 5% BurnupUncertainty, GWd/mtUPage 11 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONI Maximum k-eff 1 0.9354 0.9445 1 0.9464 0.9478Parameter Configuration 1 BEnrichment, wt% U-235 2.0 3.0 4.0 4.5Burnup credited in Calculation, 5.7 27.2 42.5 50GWd/mtUMinimum required Burnup 6.0 28.6 44.6 52.5including 5% BumupUncertainty, GWd/mtUMaximum k-eff 0.9409 0.9445 0.9465 0.9468Note that only the design basis calculations demonstrating compliance with the regulations wereupdated in the licensing report to additionally consider the effect of those aspects listed above.Other studies documented in the licensing report and in these RAI responses do not includethese revised assumptions.SummaryThe STC is a 10CFR50 component, with criticality safety requirements meeting the acceptancecriteria in 10CFR50.68. However, the design of the STC basket is based on a dry storage andtransportation design (MPC-32), which has a much higher neutron absorption than typical wetstorage systems. Also, the burnup credit methodology from a transport package design wasused, which is much more conservative (i.e. has higher minimum required burnup values) thantypical wet storage criticality methodologies. Additionally, the STC basket is flooded withborated water (2000 ppm) although it is not credited for normal conditions, and during transferoperations there is no credible event which would cause a loss of boron. Together, thiscombination results in a substantial subcriticality margin for the fuel-loaded STC in its mostreactive configuration.NRC RAI 4-1Provide a list of the isotopes and a summary table to show the biases and uncertainties that areapplied to the major actinides and the best estimate correction factors that are applied to minoractinides and fission products for which burnup credits are taken. The information provided inPage 12 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONthe revised SAR pages supplied with the previous RAI response does not adequately addressthis.This information is necessary for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-1The list of major actinides considered in the analyses is presented in Section 4.A.1.3 of thelicensing report. The minor actinides and fission products considered are listed in the supportingcriticality analysis report, HI-2084176, Table A-3. As a summary, a complete list of thoseisotopes is presented below.The bias and bias uncertainty for the major actinides is shown in the licensing report, Table4.7.14, and also shown graphically in Figures 4.A.1 and 4.A.2.It is noted that isotopic correction factors may also be available from publicly available sources,as an example see NUREG/CR-7012. However, correction factors can only be compared if theywere calculated for the same depletion code, with the same version of the code, and the samecross section library. For the example of the NUREG/CR-7012 this is not the case, since theNUREG presents correction factors for a different code, namely Triton from the Scale codesystem. Fuel compositions, if they were determined for the same enrichment, burnup, coolingtime and core operating conditions, could be determined with different depletion codes.NRC RAI 4-2Provide justification for the conclusion that 10000 histories, 80 skipped cycles, and 200accumulated cycles are acceptable for both configuration A and B, the 8 fuel assembly basket inparticular.In general, the MCNP code is sensitive to the number of cycles skipped and the ksrcspecifications for loosely coupled systems like the 8 fuel assembly basket configurations. Forcases like this, the users often have to check the convergence of both the effective neutronmultipilation factor (Keff) and the source term. The Shannon Entropy often is used to test theconvergence of the eigenvalue calculations. It is not clear, however, how these controlparameters are determined. The applicant is requested to provide justification for the selectionof the values of these control parameters.This information is necessary for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-2For MCNP4a, which does not automatically calculate the Shannon Entropy, the convergencePage 13 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONhas been previously checked by performing selected calculations with larger number of cycles,as an example see [K.C], Section 6.E.4.7. However, a newer version of the code, MCNP5, doescalculate that value. MCNP5 was therefore used to perform the evaluation of the ShannonEntropy for the three principal configurations (Configuration 1 A and 1 B for 5% enrichmentlisted in Table 4.7.1, and Configuration 2 for the 8 assembly basket with fresh fuel listed in Table4.7.2) and original and increased number of calculational cycles. The acceptance criteria fromthe MCNP5 manual were used in the evaluation, which provides a verification of the minimumnumber of skipped cycles that need to be used. Several combinations of increased number ofcycles and/or increased numbers of particles per cycle were evaluated. The evaluations showthat the total number of cycles needs to be increased by a factor of 4 while the number ofparticles per cycle is kept constant to meet the Shannon-Entropy convergence criteria. Thisrequires increasing the numbers of cycles from 200 to 800 for the configurations with spent fuel(12 assemblies), and from 160 to 640 for the fresh fuel configuration (8 assemblies). Note thatthe number of particles (histories) per cycle is kept at 10000, since studies involving changes inthis amount, with or without changes to the number of cycles did not indicate any furtherimprovements.Results are summarized in the following table, which shows the calculated k-eff for the originaland the increased number of cycles, and for MCNP4a and MCNP5. Note that the same crosssection libraries were used in both MCNP4a and MCNP5 for this comparison.Configuration 200 (Spent) / 160 (Fresh) Cycles 800 (Spent) / 640 (Fresh) CyclesMCNP4a MCNP5 MCNP4a MCNP5Spent Fuel, 0.9189 0.9189 0.9220 0.9215Configuration 1 ASpent Fuel, 0.9199 0.9190 0.9214 0.9211Configuration 1 BFresh Fuel, 0.9244 0.9228 0.9250 0.9240Configuration 2The following observations are made:* Results for MCNP4a and MCNP5 are in good agreement for each configuration,indicating that conclusions from the MCNP5 calculations are applicable to the MCNP4acalculations* The higher number of cycles results in slightly higher values in k-eff.In order to ensure that the maximum k-eff values are determined, the revised design basisMCNP calculations presented in the introduction to these RAI responses are performed with thenumber of cycles increased to 800 for spent fuel and 640 for fresh fuel. The design basiscalculation for fresh fuel shown in Table 4.7.2 of the licensing report is also re-performed, withthe new result shown below.Page 14 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONParameter 8 AssembliesEnrichment, wt% U-235 5.0Burnup, GWd/mtU 0.0Maximum k-eff 0.9357The licensing report Section 4.7.7 and the corresponding TS section 3.1.2 have been revisedaccordingly.It is noted that there is a recent publication ("Statistical Coverage Concerns in a Revised 'k-Effective of the World' Problem", Brian C. Kiedrowski and Forrest B. Brown, ANS ConferenceSummer 2011) where in addition to running a large number of cycles, the number of neutronsper cycle (NPC) needed to be increased from 10,000 to 50,000 to obtain sufficiently accurateresults. However, this was for a hypothetical configuration specifically designed to makeconvergence difficult, with features that are not present in the STC. Those features incomparison to the STC are as follows:* The geometry in the publication is a 9x9x9 array of spheres in water, with the center spherehaving a larger diameter (i.e. being more reactive). For 10,000 neutrons, only 10000/9x9x9= 14 particles would have been started initially in the one larger sphere, making it difficult todetect the specific condition of this sphere. Even for 50,000 NPC, this value increases onlyto about 70. For spent fuel (in the STC and any other basket), the more reactive sectionsare typically the top and bottom end, consisting of at least 5% of the active length at bothends. Therefore, at least 1000 particles would start in those areas even with 10,000 NPC,far more than in the center sphere with 50,000 NPC." The spheres in the geometry in the publication are loosely coupled, with a center-to-centerdistance of 20 cm and a diameter of 8 cm, i.e. a surface-to-surface distance of about 12 cm.In the STC, the dominating geometry from a criticality perspective (even for the 8 assemblybasket configuration), is the close face-to-face positioning of the assemblies that createsintensive coupling between the assemblies.* The geometry in the publication contains asymmetric coupling through a cadmium layer onthe inner (larger) sphere. No such strong asymmetry exists in the STC, although there wouldbe some small asymmetry due to the fact that the neutron absorber plates are only locatedon one side of every basket wall.In summary, the geometry in the STC is different from the relevant features in the publicationthat result in the convergence problem. The number of 10,000 NPC derived from the evaluationof the Shannon Entropy is therefore considered sufficient for the calculations for the STC.NRC RAI 4-3Provide justification for the applicability of the selected critical experiments to the codebenchmark for the system of the eight fuel assembly configuration.Page 15 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONFrom the selected benchmark critical experiments, it appears that none of them has aconfiguration similar to the configuration of the eight fuel assembly basket. The applicant isrequested to provide justification for the applicability of the selected critical experiments for theeight fuel assembly basket.This information is necessary for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-3The benchmark experiments do not have a configuration similar to the 8 assembly basketconfiguration, where a large water space exists in the center of the basket. However, this largewater gap is not the dominating feature from a criticality perspective since there is practically noneutronic interaction across such a large water gap of more than 18 inches. The dominatingconfiguration from a criticality perspective is therefore the close location of several assemblies,which is well represented by the standard benchmark experiments. The critical experiments thatwere used are therefore sufficient, and no further critical experiments are required for thisconfiguration. Further note that the results for the 8 assembly basket configuration showsubstantial margin to offset any small differences in bias and bias uncertainty.NRC RAI 4-4Review and provide a corrected term of relative burnup in reference to Table 4.5.3.Table 4.5.3 contains two pairs of columns, one pair is relative burnup at zero burnup and thesecond pair is relative burnup at 45 gigawatt days per metric ton uranium (GWD/MTU). Itappears that the term "relative burnup" should be "normalized power." The term "relativeburnup" is not determined if there is no burnup; 0/0 is not defined. The applicant is requested toreview and provide corrected information as necessary to accurately represent the data.This information is necessary for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-4The intent of Table 4.5.3 is to specify two data points for each axial node to allow the calculationof the relative burnup for any given assembly average burnup, but not to imply that there is anaxial burnup distribution for an assembly with zero burnup. The term "relative burnup" istherefore correct. Table 4.5.3 has been changed to the version below, where the zero burnupcolumn is replaced by a 5 GWd/mtU column, with appropriate relative burnups to represent thesame linear relationship between assembly average burnup and relative burnup for each nodeas before. Additionally, changes and additions were made to the data in the table as follows:Initially, generic, though conservative, axial burnup profiles were used. However, IP3 profilesare now available for three recent cycles (Cycles 14 through 16) of unloaded fuel, for a totalof about 290 fuel assemblies. These profiles are for fuel with enriched axial blankets. Asingle axial profile has been determined that bounds the relative burnup profile in all thoseassemblies, regardless of the burnup. This profile is also listed below, together with thePage 16 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONcorresponding axial segmentation, which is non-uniform in the areas near the ends of theassembly.The evaluation for the Westinghouse 15x1 5 non-blanketed assemblies was based on adataset of only 4 assemblies, over a narrow burnup range. Determination of linear functionsof the relative burnup in each node may therefore not be appropriate. Instead of thislinearization, all four individual burnup profiles are shown and evaluated.For a discussion on how those profiles are used please see response to RAI 4-18.Table 4.5.3Axial Burnup DistributionNRC RAI 4-5Provide a justification of the burnup credit calculations submitted, or resubmit with lowerbounding specific power.In Section 4.7.1.2.1 of the SAR, the applicant states that the maximum value for all plants withWestinghouse 15x15 fuel assemblies is used as bounding assembly specific power. However,studies published in "NUREG/CR-6665, Review and Prioritization of Technical Issues Related toBurnup Credit for LWR Fuel" and "M. D. DeHart, Sensitivity and Parametric Evaluations ofSignificant Aspects of Burnup Credit for PWR Spent Fuel Packages, ORNITM-12973,Lockheed Martin Energy Research Corp., Oak Ridge National Laboratory, May 1996" indicatethat using lower specific power produces conservative results for casks taking burnup credit forboth actinides and fission products. To ensure criticality safety, lower specific power should beused for applications that take credit for both actinides and fission products. The applicant isrequested to review and redo its burnup credit analyses using appropriate specific power.This information is necessary for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-5While the effect of the specific power on the maximum k-eff is comparatively small (see studiespresented in Table 4.7.16 of the licensing report), it is recognized that a lower specific power,not a higher specific power, is the more conservative assumption, as discussed in NUREG/CR-6665. To take this into consideration, the revised design basis analyses have been performedwith a conservatively low specific power. A specific power of 60% of the core average specificpower was used as the conservatively low value. This assumption is based on the studiesPage 17 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONperformed for the HI-STAR 100 Methodology, documented in Appendix A of [L.K], whichindicate that only a small fraction of assemblies (less than 1%) would have a discharge burnup(and hence average specific power) of less than 60% of the corresponding core average value,and is well below the lowest specific power for any assembly at IP3.Section 4.7.1.2.1 of the licensing report has been updated to incorporate the above discussion.Note that this change presents a slight deviation from the original HI-STAR 100 methodology; itis more conservative.NRC RAI 4-6Demonstrate that the fuel assembly depletion code, CASMO-4, is capable of modeling both thepoisoned portion and the non-poisoned portion of the wet annular burnable absorber (WABA)rods in a fuel assembly.Page 4-25 of the revised SAR states that the length of the poisoned region of WABA rods variesfrom 120 to 128 inches. The top and bottom parts of a WABA rod are just cladding. This leavesthe top and bottom parts of a fuel assembly unexposed to poison. The SAR further states:"Most of the calculations with WABAs take that into consideration, i.e., the depletion calculationsfor the top and bottom node are performed with the cladding of the WABA in place, but withoutany poison while the depletion calculations for the central part of the assembly contains both theWABA cladding and the absorber." It is not clear, however, how the CASMO-4 code modelsthis three dimensional effect for the fuel assembly containing WABAs. The applicant isrequested to demonstrate that the fuel assembly depletion code, CASMO-4, is capable ofmodeling both the poisoned portion and the non-poisoned portion of the WABA rods in a fuelassembly.This information is necessary for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-6To alleviate any concerns regarding the modeled length of the burnable poison rods, and tomake the methodology consistent with that used for the HI-STAR 100, the revised design basiscalculations have been modified such that the burnable poison rods and IFBA coatings nowcover the entire active length of the fuel.NRC RAI 4-7Demonstrate that reactivity effect of the burnup gradient is negligible for the 12 fuel assemblyloading configuration.Page 4-29 of the revised SAR discusses the potential reactivity impact of burnup gradientsacross the fuel assemblies and further states: "However, since this is a highly unlikelyconfiguration, and since even then the reactivity is small compared to the remaining safetymargin (see Section 4.7.9), all design basis calculations are performed with a uniform planarburnup distribution." This assessment, however, may not be justified without a quantitativeanalysis of the system with consideration of the burnup gradient in the fuel assemblies. Thestaff is particularly concerned with the small system 12 fuel assembly shielded transfer cask.The applicant is requested to demonstrate that reactivity effect of the burnup gradient isnegligible for the 12 fuel assembly loading configuration.Page 18 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONThis information is necessary for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-7The effect of burnup gradients has now been evaluated, using the same approach that wasutilized in [K.C]. For this evaluation the bounding burnup gradient is assumed in each assembly,in addition to assuming that the lower burned section of each assembly is faced inwards towardthe center of the basket. This is a simple but very conservative assumption, creating aconfiguration that is highly unlikely in reality. Results are listed in the table below, together withresults of the corresponding calculations that do not consider any burnup gradients.Axial Burnup profile ConstantEnrichment, wt% 2 3.5 5Burnup, GWd/mtU 10 30 45K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calcWABA -Reference 0.8751 -0.8916 -0.9098 -WABA -Tilted Burnup 0.8766 0.0015 0.8952 0.0036 0.9144 0.0046RCCA -Reference 0.8985 -0.9294 -0.9479RCCA -Tilted Burnup 0.8974 -0.0011 0.9324 0.0030 0.9508 0.0029Axial Bumup Profile ProfileEnrichment, wt% 2 3.5 5Burnup, GWd/mtU 10 30 45K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calcWABA -Reference 0.8754 -0.9041 -0.9169 -WABA -Tilted Bumup 0.8765 0.0011 0.9099 0.0058 0.9212 0.0043RCCA -Reference 0.8960 -0.9322 -0.9471RCCA -Tilted Burnup 0.8964 0.0004 0.9365 0.0043 0.9497 0.0026The results show reactivity effects of up to 0.0058 delta-k. This is slightly larger than the effecttaken from [K.C], which was 0.0038.To alleviate any concerns regarding the radial burnup gradients, the revised design basiscalculations assume radial burnup gradients in all analyses. Note that this change presents aslight deviation from the original HI-STAR 100 methodology; it is more conservative.NRC RAI 4-8Demonstrate that the chemical assay data used for code benchmarking are sufficient to coverfuel assemblies with burnup exceeding 50 GWD/MTU.Page 4-32 of the revised SAR states: "Note that ISG-8 Rev. 2 recommends an upper limit forburnup credit of 50 GWD/MTU, based on an apparent lack of data above this value. However,the evaluation of the bias and bias uncertainties and isotopic correction factors fromPage 19 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONbenchmarking have been performed using a statistical approach that accounts for the limitedamount of data available and assigns higher uncertainties for parameters that are further awayfrom (i.e. at higher burnups) the experimental data. Also, isotopic benchmarking includes datafor fuel exceeding 50 GWD/MTU. An additional burnup limit does therefore not appear to benecessary and is therefore not applied here." The staff reviewed the relevant publications andfound that the bias and bias uncertainties determined based on the comparisons of thecalculated and experimental isotopic concentrations are valid only for the burnup range of theexperimental data. Based on the staff's review, it appears that there is a very limited number ofchemical assay data that have burnup in excess of 50 GWD/MTU and the statistical analysesusing the very limited data points may not satisfy the criterion for obtaining meaningful results.In addition, it appears that the burnup values are the local burnup of samples rather than thefuel assembly average burnup that is used in fuel qualification calculation. The applicant isrequested to demonstrate that the chemical assay data used for code benchmarking aresufficient to cover fuel assemblies with burnup exceeding 50 GWD/MTU. If the criticalityanalyses are limited to 50 GWD/MTU, revise TS LCO 3.1.2.a.2, 3.1.2.b.2, and Note b to TSTable 3.1.2-1 to show the revised limit.This information is necessary for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-8Entergy understands that this aspect of the STC evaluation did not follow the approachdeveloped and presented for the HI-STAR 100 in [K.C]. To bring the STC analysis in line withthe HI-STAR 100 analysis, the credit for burnup is now limited to 50 GWd/mtU. This only affectsConfiguration 1 B, since only for Configuration 1 B burnup requirements above 50 GWd/mtUwere initially calculated. For this configuration, a maximum enrichment of 4.5 wt% is now used,with a credit of 50 GWd/mtU (fuel above this enrichment, will need to be transferred using the 8assembly configuration that does not credit burnup). Note that to account for a burnup recorduncertainty of 5%, the corresponding minimum required fuel burnup may be as high as 52.5GWd/mtU, however, the amount of burnup credited is still only 50 GWd/mtU, consistent withISG-8 Rev. 2.NRC RAI 4-9[deleted]NRC RAI 4-10Page 4-35 of revision 3 of the licensing report makes reference to Table 4.7.11 and explains thecases evaluated in that table. To clarify the change between the second to last and last rowsplease see Table 4.7.1. This table shows the combination of bias and bias uncertainty for alldesign basis calculations. It is important to note there that the bias for various conditions islisted as a negative value, i.e. one that would reduce the maximum k-eff if considered. Thisreduction in not applied in the design basis calculation, consistent with guides such asNUREG/CR-6698. Biases are therefore truncated to 0 if they are below 0 in Table 4.7.1However, in the second-to-last row of Table 4.7.11, the negative biases are applied, i.e. theywould offset some uncertainties and reduce the maximum k-eff. Note that this the negative biasis only applied to the studies presented in Table 4.7.11 that evaluate margin, not for the designbasis calculation, and there is no claim that such an approach would be acceptable for designbasis calculations that determine loading curves. Therefore, going from the second-to-last rowPage 20 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONto the last row in Table 4.7.11, all biases that would reduce maximum k-eff are set to zero, sothe last row represents the design basis calculation as discussed in the text on Page 4-35.Response to RAI 4-10Page 4-35 of revision 3 of the licensing report makes reference to Table 4.7.11 and explains thecases evaluated in that table. To clarify the change between the second to last and last rowsplease see Table 4.7.1. This table shows the combination of bias and bias uncertainty for alldesign basis calculations. It is important to note there that the bias for various conditions islisted as a negative value, i.e. one that would reduce the maximum k-eff if considered. Thisreduction is not applied in the design basis calculation, consistent with guides such asNUREG/CR-6698. Biases are therefore truncated to 0 if they are below 0 in Table 4.7.1.However, in the second-to-last row of Table 4.7.11, the negative biases are applied, i.e. theywould offset some uncertainties and reduce the maximum k-eff. Note that this negative bias isonly applied to the studies presented in Table 4.7.11 that evaluate margin, not for the designbasis calculation, and there is no claim that such an approach would be acceptable for designbasis calculations that determine loading curves. Therefore, going from the second-to-last rowto the last row in Table 4.7.11, all biases that would reduce maximum k-eff are set to zero, sothe last row represents the design basis calculation as discussed in the text on Page 4-35.NRC RAI 4-11[deleted]NRC RAI 4-12Provide detailed information on and justification for the interpolation scheme used indetermining the isotopic concentrations as a function of burnup.Page 4-30 of the revised SAR states: "Since it is necessary to model the axial burnupdistribution, a large number of isotopic compositions at irregular burnups are required. Given thesignificant number of criticality calculations and studies performed for the burnup creditevaluations, it would be impractical to perform CASMO-4 depletion calculations for each ofthese burnups. Instead, CASMO-4 runs are performed for fixed burnups at 2.5 GWD/MTUintervals (or less), and intermediate isotopic values are determined by linear-linear interpolation.Calculation presented in [16] show that this is an acceptable approach by comparing resultswith calculations that were based on a smaller interval of 1.0 GWD/MTU in the CASMOcalculation." From these statements, it is not clear what interpolation scheme was used indetermining the isotopic concentrations at various burnup values. It is not clear whether theisotopic concentrations of the various isotopes of interest can be determined with thisinterpolation scheme. The applicant is requested to provide detailed information on andjustification for the interpolation scheme used in determining the isotopic concentrations as afunction of burnup.This information is necessary for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-12The interpolation scheme is linear-linear (see Section 4.7.3.1 of the licensing report), i.e. thePage 21 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONconcentration of each isotope is assumed to be a linear function of burnup within each burnupinterval. In the original development of the methodology, additional calculations with a 1GWd/mtU interval had been performed to justify the interpolation scheme, as recognized in theRAI. To show that this conclusion is also applicable to the STC, those studies were re-performed for the STC, and results are listed below together with current results for 2.5GWd/mtU intervals.Axial Bumup profile ConstantEnrichment, wt% 2 3.5 5Bumup, GWd/MTU 10 30 45K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calcWABA -Reference 0.8751 -0.8916 -0.9098WABA -1 GWD 0.8752 0.0001 0.8901 -0.0015 0.9097 -0.0001RCCA -Reference 0.8985 -0.9294 -0.9479 -RCCA -1 GWD 0.8988 0.0003 0.9287 -0.0007 0.9473 -0.0006Axial Bumup Profile ProfileEnrichment, wt% 2 3.5 5Bumup, GWd/MTU 10 30 45K-calc Delta k-calc K-calc Delta k-cal K-calc Delta k-calcWABA -Reference 0.8754 -0.9041 -0.9169 -WABA -1 GWD 0.8753 -0.0001 0.9044 0.0003 0.9167 -0.0002RCCA -Reference 0.8960 -0.9322 -0.9471RCCA -1 GWD 0.8963 0.0003 0.9332 0.0010 0.9466 -0.0005The statistical uncertainty is about 0.0015 delta-k (95/95) for all delta-k values. As can be seen,all delta-k values are at or below that value, indicating that the differences are statisticallyinsignificant. The current interpolation scheme using 2.5 GWd/mtU is therefore acceptable.Note that the design basis calculations are always performed with burnable absorbers or controlrods inserted for the entire irradiation period of the assembly. There are therefore no situationswith possible discontinuities in isotopic concentration from the insertion or removal of rods thatmay require an adjustment of the interpolation scheme.NRC RAI 4-13Provide a list of isotopes included in the baseline cases presented in page 4-35 and Table4.7.11 and justification for including all isotopes in the baseline case.Page 4-35 of the revised SAR presents a baseline scenario and comparisons with the results ofvarious ways of treating the uncertainties involved in the isotopic concentrations of spent fuelscalculated using the CASMO code. From the description in the text and the results presented inTable 4.7.11 of the SAR, it seems that the applicant indicates that all isotopes that are trackedby the CASMO code were included in the STC criticality calculations. From page C-6 ofAppendix C to BURNUP CREDIT FOR THE MPC-32, Holtec Report No. HI-2012630, it appearsthat Kr-83 was included in the burnup credit analysis. Page C-64 of Appendix C to BURNUPCREDIT FOR THE MPC-32, Holtec Report No: HI-2012630, further states: "Two series werePage 22 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONperformed, one for an enrichment of 4 percent at a burnup of 32.5 GWD/MTU, and one for anenrichment of 5 percent and at a burnup of 50 GWD/MTU. The reference calculations use allisotopes from CASMO without any corrections factors, including the lumped fission products."From these statements or input files in the various documents, it was not clear what exactlythose isotopes are and what the justification is for including each of those isotopes. It is thestaff's understanding that some of the isotopes, such as krypton and xenon, cannot be creditedbecause these isotopes are gaseous at the expected temperature. Because the exact locationsof these isotopes in the fuel rods cannot be determined, it is not considered justified to includedin criticality calculations, even in the Commercial Reactor Critical experiments. The applicant isrequested to provide a list of all the isotopes that were included in the baseline calculation andjustification for inclusion of each of the isotopes in the criticality calculations.This information is necessary for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-13The text in the Holtec Report HI-2012630 referred to in the RAI belongs to earlier calculationsthat used a larger set of isotopes. The final design basis calculations in [K.C] were documentedin Supplement 2 of report HI-2012630. Those calculations, as well as the design basiscalculations for the STC, use only the isotopes listed in the table provided in the response toRAI 4-1. Specifically, those calculations do not credit isotopes such as krypton, xenon, orlumped fission products. The only isotopes credited are those that are validated through thechemical assays. However, note that the full set of isotopes available from CASMO is utilized inthe calculations listed in the first row in Table 4.7.11 to estimate the margin of the analysis.The full set of isotopes is also used in the CRC calculations. This is necessary since only a bestestimate approach (using all isotopes) in a benchmark calculation can provide a meaningful biasand bias uncertainty. This is recognized in NRC's SER for the HI-STAR 100 (ML062860201),which states on Page 22 "The CRC benchmark calculations included 179 isotopes, whereas,the design calculations included 25 isotopes. This approach is consistent with the principle thata best estimate which includes all known effects that reduce kea should be used whenestablishing the bias of the criticality calculations even if these effects are not used in the designcalculations". In this context please see also the discussion on the relevance of the CRCbenchmarks for the safety analysis, as discussed in the response to RAI 4-16.NRC RAI 4-14Revise the context of the discussions regarding the temperature impact to neutron fluxes atCRCs and MPC on page A-25 of the Appendix A to the Holtec Report No. Hi-2084176 to makethe context consistent with the referenced figures.Page A-25 of the Appendix A to Holtec Report No. Hi-2084176 discusses the temperatureimpact to neutron fluxes at CRCs and MPC. However, it appears that there are no differencesbetween Figure 53 and Figure 53a. The discussions on page A-25 of the report does not makereference to Figure 53a either. The applicant is requested to revise the context of thediscussions regarding the temperature impact to neutron fluxes at CRCs and MPC on page A-25 of the Appendix A to the Holtec Report Hi-2084176 to make the context consistent with thereferenced figures. More importantly, the applicant should explain why there is no spectraldifference between these two systems given the fact that the Doppler Effect at 600K issignificant different from that at 300K.Page 23 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONThis information is necessary for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-14It is correct that there is no difference in the data presented in Figure A.53 and Figure A.53a.Figure A.53a simply shows the same information as Figure A.53, just with a linear y-axis. Thetext contains a typographical error as follows: the sentence "Figure A.53 shows the same dataon a linear plot." should have stated "Figure A.53a shows the same data on a linear plot." TheHoltec Report No. HI-2084176 has been updated to correct this.The reason that there is essentially no spectral difference between the two systems (CRC andMPC) in those plots is that for those plots, the MPC was also evaluated at 600K, which is thesame temperature used in the CRC calculations. As discussed on Page A-25, this is to showthat given the same temperature, the spectra between CRCs and MPCs are essentially thesame. The discussion on Page A-24, together with Figures A.50 through A.52a, addresses thecomparison of spectra at different temperatures. Further discussion on the temperature is alsoincluded on Page A-25. In this context please see also the discussion on the relevance of theCRC benchmarks for the safety analysis, as discussed in the response to RAI 4-16.NRC RAI 4-15Explain what has been done in the benchmark calculations to account for the differences in theneutron absorption rates between the commercial reactor criticals and the MPC.Pages A-22 and A-25 of Appendix A of the Holtec Report No. Hi-2084176 provides detailedcomparison of the B-1 0 reaction rates in the Commercial Reactor Criticals and MPC. Fromfigure A.55a or figure A.55b, it appears that the difference is fairly significant and the applicantrecognized this difference in the discussion. However, it was not clear what has been done inthe benchmark calculation to account for the differences. The applicant is requested to explainwhat has been done in the benchmark calculations to account for the differences in the neutronabsorption rates between the commercial reactor criticals and the MPC.This information is necessary for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-15The studies on B-10 discussed on Pages A-22 through A-25 of Holtec Report HI-2084176 wereoriginally performed during the HI-STAR licensing process since there had been the concernthat the reactivity effect of the B-1 0 would be substantially different between the CRCs and theMPC. However, the results presented in Figure A.35 indicate that the reactivity effect of B-10 inthe CRC and MPC are in fact comparable. This alleviated the concerns about the representationof B-10 in those benchmarks. Therefore, while an additional comparison of reaction rates wasalso performed (presented in Figures A.55a and A.55b), no further actions resulted from thisstudy in the HI-STAR 100 burnup credit methodology. In this context, please also see theresponse to RAI 4-16 below on the purpose and relevance of the CRC benchmarks for thesafety analyses.NRC RAI 4-16Page 24 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONDemonstrate that the CRC benchmark calculations could be used as the sole benchmarking forthe burnup credit methodology.On pages A-26 of Appendix A to Holtec Report No. Hi-2084176, the applicant concluded, aftersome discussions, "Based on these results and conclusions, the CRC benchmark calculationscould justifiably be used as the sole benchmarking for the burnup credit methodology. The onlybias would then be the uncertainty of the CRCs (when conservatively ignoring the negativebias), and this would support the use of all isotopes generated by CASMO, without anycorrection factors." Given the fact that the CRC models that the applicant built included manyisotopes that are not available to bumup credit ( see RAI 4-13), if not carefully considered somesignificant errors might have been introduced in the benchmark calculations using the CRCs.Although the results of code benchmark with inclusion of all isotopes CRC models show goodresults, using all of the isotopes generated by CASMO may unintentionally affect the accuracyof the bias and bias uncertainty. The applicant is requested to demonstrate by calculations andanalyses that all isotopes from the CASMO calculations are available and appropriate forburnup credit. Quantitative calculations are necessary to quantify the bias and biasuncertainties associated with each isotope to be included in the burnup credit analyses.This information is necessary for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-16The CRC benchmarks were not used as the sole benchmark as explained in the text followingthe quoted section from Page A-26. The text continues "However, as a more conservativeapproach, and more consistent with the recommendations in ISG 8, the primary benchmarkexperiments to support the burnup credit methodology are selected to be the criticalitybenchmarks based on fresh U02 and MOX fuel, and the isotopic benchmarks of spent fuel. TheCRC benchmarks are used only to provide additional support for credit of the fission productsand minor actinides, specifically regarding the uncertainty in the cross sections of theseisotopes." This is reflected in NRC's SER for the HI-STAR 100 (ML062860201), which states onPage 22 "Staff determined that the CRCs can provide some useful information when coupledwith other benchmark data but does not believe that the applicability of the CRCs has beendemonstrated to the degree that they can be used as the sole means of benchmarking thecriticality calculations for the analysis of the MPC-32 package for transport of spent fuel.", butthen continues "Based on the margin of safety and on risk informed considerations of theprobability and expected consequences to public health and safety from the use of the MPC-32for transport, staff accepts the CRCs as part of this overall benchmark analysis.". Using theCRCs only to support credit for selected minor actinides and fission products significantlyreduces the overall importance of the CRC benchmarks on the safety analyses, compared tocalculations that would be intended as sole benchmarks. Nevertheless, the full uncertaintyderived from the CRCs is applied for the design basis safety analyses, and is combined with theuncertainties and adjustments from the other benchmarks.It is certainly true that a more desirable situation from a validation perspective would be to havea bias and bias uncertainty for each isotope and for both depletion analyses (i.e. concentrationof the isotope) and criticality analysis (i.e. reactivity worth of the isotope). Unfortunately, thecurrent state of the available benchmarking experiments does not fully support such anapproach: We have chemical assays for actinides and fission products in spent fuel and onlyuse those isotopes validated through those experiments, with appropriate correction factorsPage 25 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATION(see also response to RAI 4-1). We also have the validation of the criticality calculations ofmajor actinides through the fresh and MOX fuel critical. However, there are no significantpublicly available experiments to validate the reactivity worth of individual minor actinides andfission products. For this very reason, the CRCs were analyzed as additional benchmarks andwith NRC concurrence (see excerpts from NRC's SER listed earlier in this response), since theyat least evaluate the combined effect of those isotopes. While this is not the perfect validation, itwas considered sufficient to support the safety evaluation of the HI-STAR system.In order to determine any bias and bias uncertainty from the CRC benchmarks, it is in factnecessary to perform those as best estimate calculations and include all isotopes available inCASMO. As already discussed in the response to RAI 4-13, this is recognized in NRC's SER forthe HI-STAR 100 (ML062860201), which states on Page 22 "The CRC benchmark calculationsincluded 179 isotopes, whereas, the design calculations included 25 isotopes. This approach isconsistent with the principle that a best estimate which includes all known effects that reduce kIfshould be used when establishing the bias of the criticality calculations even if these effects arenot used in the design calculations".NRC RAI 4-17If applicable, review and correct the reference on page 11 of Holtec Report No: HI-2094486,"MCNP Benchmark Calculations."Page 11 of Holtec Report No: HI-2094486, "MCNP Benchmark Calculations," cites reference 5in the reference list of this report. There appears to be an error in this reference becausereference 5 of the report is NUREG/CR-6998 while the text of the report seems to refer to anerror analysis technique for normality test of data used in statistical analyses. The applicant isrequested to review the document and make corrections if necessary.Response to RAI 4-17No corrections are necessary. The Shapiro-Wilk test and the corresponding tables are listed inNUREG/CR-6698 (not 6998), which is reference [5] in HI-2094486.NRC RAI 4-18Regarding the burnup profile used in the burnup credit analyses of the Indian Point fuel transfershielded transfer canister:* Explain why the burnup profile derived in Appendix D to Holtec report HI-2012630 is notconsistent with that developed in NUREG/CR-6801;* Explain why the burnup profiles, as shown on pages D-19 to D-23, are not normalized;" Explain why the burnup profiles, as shown on pages D-35 and D-36, have the divergedsections and justification for their applicability to the STC;* Provide justification for the applicability of the burnup profile for future discharged fuelassemblies.This information is needed for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-18Page 26 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONBefore the four sub-questions of this RAI are addressed, a brief discussion on axial profiles(enrichment and burnup) is presented, together with the comparisons of the effects of profileson the design basis calculations.Enrichment ProfilesAs at many other plants, Indian Point 3 has initially used fuel with an axially constantenrichment, followed by using fuel with natural uranium axial blankets, and finally using fuel withenriched (2.6% and 3.2%) axial blankets. Fuel with different axial enrichment profiles also showdifferent burnup profiles, with lower relative burnups in the blanket sections. Previouscomparisons of fuel with different enrichment profiles have consistently shown that given thesame assembly average burnup, blanketed fuel (both natural and enriched blankets) is boundedby (i.e. less reactive than) fuel with constant enrichment over the entire active height. This isdue to the reduced U-235 amount in the blanket areas that only have a low burnup. Othercriticality analyses, such as the recently approved Beaver Valley analysis, have credited thisfact by developing different loading curves for different enrichment profiles, with lower burnuprequirements for blanketed fuel. For IP3 fuel, no such approach is used, and the burnuprequirements are based on the most reactive, i.e. axially constant, enrichment profile.Nevertheless, various enrichment and burnup profiles are evaluated for comparison purposesand to show that the approach is conservative.Burnup profiles used in the AnalysesInitially, only generic burnup profiles were used: In the initial calculation that used a typical wetstorage burnup credit methodology, profiles from NUREG/CR-6801 were used, whereas later,the profiles developed for Westinghouse 17x17 assemblies with the HI-STAR 100 methodologywere used when the HI-STAR 100 methodology was used for the STC, supplemented withadditional profiles from the same source used in the NUREG (i.e. the Profile Database fromYankee Atomic Report YAEC-1 937). Burnup profiles from discharged fuel at IP3 are nowavailable. However, those are only for three recently discharged cycles (Cycles 14 through 16),and for fuel with enriched blankets. To qualify all currently discharged fuel, and provide a basisfor qualifying fuel discharged in the future, calculations have been performed with all applicableaxial burnup and enrichment distributions. These distributions are discussed below, with aspecific focus on applicability and inherent conservatisms:1. Profiles for Westinghouse 17x1 7 assemblies without any axial blankets, developed forthe HI-STAR 100 burnup credit methodology. These are the bounding profiles for mostconditions, i.e. they result in the highest maximum k-eff. Since assemblies without axialblankets were only used in earlier cycles, they have a much longer cooling time than the5 years assumed in the design basis criticality analysis, which provides additional butunspecified margin.2. The profile database that was used to develop the Westinghouse 17x1 7 profiles for theHI-STAR 100 only contained 4 profiles for Westinghouse 15x15 assemblies, which is thedesign used at IP3. All four profiles were used, to confirm that results for those 15x15profiles are equivalent to or bounded by results obtained for the 17x1 7 profiles.3. Profiles from NUREG/CR-6801 are used, to provide further confirmation that the profilesfrom the HI-STAR 100 methodology are appropriate.Page 27 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATION4. An axially constant burnup and enrichment distribution. While this distribution does notrepresent any realistic condition, it is traditionally used for added conservatism since ittypically result in higher k-eff values at lower enrichments and burnups than the actualprofiles.5. Profiles for Westinghouse 15x1 5 and 17x1 7 assemblies with natural blankets, also fromthe Profile Database form Yankee Atomic Report YAEC-1 937.6. Profiles from IP3 for recently discharged fuel (Cycles 14 through 16) with enrichedblankets.Results for the calculations with the various profiles are summarized in the following tables.Parameter Configuration I AEnrichment, wt% U-235 2 3 4 5Burnup, GWd/mtU 5.1 21 35.1 46.4Max. k-eff Delta-k Max. k-efi Delta-k Max. k-ef] Delta-k Max. k-efi Delta-k-Reference profile -W17x17 0.9349 -0.9445 -0.9464 -0.9478 NUREG 0.9339 -0.0010 0.9386 -0.0059 0.9427 -0.0037 0.9468 -0.00104- Flat 0.9354 0.0005 0.9325 -0.0120 0.9296 -0.0168 0.9362 -0.01165 -Natural Blankets --0.9244 -0.0201 0.9170 -0.0294 0.9208 ]"-0.02706- Enriched Blankets .- -0.9301 -0.0163 0.9370 1 -0.0108Maximum k-eff 0.9354 0.9445 0.9464 0.9478Profile with Max. k-efl Flat W17xl7 W17x17 W17xl7..... ... .... ........... .. ..... ..... .... ..... ....... ...Parameter Configuration I BEnrichment, wt% U-235 2 3 4 4.5Burnup, GWd/mtU 5.7 27.2 42.5 50Max. k-eff Delta-k Max. k-efl Delta-k Max. k-efl Delta-k Max. k-ef Delta-k-Reference profile -Wl7xl7 0.9409 -0.9445 -0.9465 -0.9468 NUREG 0.9400 -0.0009 0.9423 -0.0022 0.9428 -0.0037 0.9450 -0.00184-Flat 0.9409 0.0000 0.9425 -0.0020 0.9446 -0.0019 0.9456 -0.00125-Natural Blanktes --0.9349 -0.0096 0.9334 -0.0131 0.9335 -0.01336 -Enriched Blanktes ---0.9421 r -0.0044 0.9414 1 -0.0054Maximum k-eff 0.9409 0.9445 0.9465 0.9468Profile with Max. k-eff Wl7x17 W17xl7 W17x17 Wl7xl7Page 28 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONParameter Configuration I AAssembly No. 1 2 3 4Enrichment, wt% U-235 4 4 4 4Burnup, GWd/mtU 35.1 35.1 35.1 35.11 -Reference k-eff- W17x17 0.94642 -W15x15 0.9283 0.9333 0.9320 0.9414Delta k-eff -0.0181 -0.0131 -0.0144 -0.0050Parameter Configuration 1 BAssembly No. 1 2 3 4Enrichment, wt% U-235 4 4 4 4Burnup, GWd/mtU 42.5 42.5 42.5 42.51 -Reference k-eff- W17x17 0.94652 -W15x15 0.9408 0.9431 0.9415 0.9477Delta k-eff -0.0057 -0.0034 -0.0050 0.0012The results support the following conclusions:" The non-blanketed profiles (Westinghouse 15x15, 17x17, Flat) are always bounding. Asstated above, this provides additional margin since those assemblies have a longercooling time than the 5 years used in the design basis criticality calculations." The Westinghouse 15x15 profiles and the profiles from the NUREG result in similar k-effvalues as, or are bounded by the Westinghouse 17x17 profiles, providing additionalassurance that the selection of the axial profiles is appropriate and conservative." Profiles for fuel with natural and enriched blankets result in maximum k-eff values thatare significantly lower than those for the non-blanketed assemblies. The difference isbetween about 0.01 and 0.03 delta-k for natural blankets, and between about 0.005 and0.015 delta-k for enriched blankets. This provides additional margin for those assemblytypes, which may have cooling times closer to 5 years used in the analysis.Overall, the analysis is considered adequate and conservative for all fuel used at IP3, and allfuel assemblies currently in the IP3 spent fuel pool, up to and including assemblies unloaded incycle 16. Future assemblies, i.e. assemblies unloaded from cycle 17 and following, will beevaluated before they are loaded into the STC to ensure they are bounded by the design basisanalyses.Responses to the individual questions in the RAI are listed below.1. HI-2012630 vs. NUREG/CR-6801:The axial burnup profile approach for the HI-STAR 100 was developed in parallel toNUREG/CR-6801, but both are essentially based on the same profile database, namely thedatabase documented in the Yankee Atomic Report YAEC-1937, which was also used in anearlier report from Texas A&M University, "Bounding Axial Profile Analysis for the TopicalReport Database" (Ref. 7 in the NUREG). The main technical reasons for developing aPage 29 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONmethodology different from NUREG/CR-6801 and the Texas A&M University report in the HI-STAR 100 burnup credit methodology are as follows:" The NUREG and the Texas A&M University report base the comparison of the profiles onone-dimensional (1-D) calculations. Holtec developed bounding profiles directly from theprofile database. This is believed to be a more rigorous approach since it avoids the relianceon the 1-D calculations.* The profiles listed in the NUREG and the Texas A&M University report create steps in theloading curves, which make application difficult. The Holtec-developed profiles arecontinuous functions of burnup in each axial section.To alleviate any concerns with respect to the selection of burnup profiles, additional calculationswere added with the profiles from NUREG/CR-6801. Results and comparisons to thosecalculations that use the profiles developed for the HI-STAR 100 methodology are presentedearlier in this response.2. Normalization:A lower bound relative burnup as a function of the assembly average burnup is determinedseparately for each axial node. This results in non-normalized profiles. Since the value in eachaxial section is a lower bound value, the average over all sections in any given profile willalways be below 1.0. However, the linearizations for the non-blanketed 15x15 assemblies wereonly performed from a small dataset of 4 assemblies over a small burnup range, resulting inaverage values that exceeded 1.0 for higher burnups. The linearization has therefore beenremoved, and the four profiles are individually analyzed (see also response to RAI 4-4).3. Diverged Sections on Pages D-35 and D-36:These show the effect of axial power shaping rods present in the active region during full poweroperation, an approach used in B&W reactors. As a result, the methodology in [K.C]distinguishes between B&W and Westinghouse assemblies in terms of bounding axial burnupprofiles, and the profiles for B&W fuel show the effect of those axial power shaping rods whilethe profiles for Westinghouse fuel do not. Note that IP3 is a Westinghouse reactor and has notused axial power shaping rods, except in the first cycle, i.e. in fuel that has a significant coolingtime. However, the profiles developed in NUREG/CR-6801 are now used in the analyses inaddition to the profiles developed for the HI-STAR 100, and those do include all profiles from thedatabase.4. Future Profiles:Procedures will be implemented to screen axial burnup distributions for future cycles to ensurethat future fuel transferred with the STC will be bounded by the design basis criticalitycalculations for spent fuel. If it cannot be demonstrated that the spent fuel is bounded by thedesign basis calculations, then the fuel assembly will be transferred in the 8 assemblyconfiguration that does not credit fuel burnup. This applies to all fuel discharged from cycle 17and following, since Cycle 16 is the last cycle whose profiles are already included in the currentevaluation.Note that this represents a deviation from the HI-STAR approach, where no additional screeningof axial distributions is required. However, it is recognized that this screening requirement isPage 30 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONconsistent with recently approved 1 OCFR50 criticality analyses such as that for Beaver ValleyUnit 2 (ML1 10890844, SER Section 3.2.4).NRC RAI 4-19Provide a justification for the conclusion made on page 2 of Holtec Report HI-2032973 that theCRC benchmarks tend to overestimate the actual reactivity and demonstrate that the modelsbuilt for code benchmarking did not introduce additional bias in the criticality safety analyses.Page 2 of Holtec Report, "Commercial Reactor Critical Benchmarks for Burnup Credit," states:"The average of the calculated reactivities is 1.0023, i.e., the calculations tend to overestimatethe actual reactivity...." It is not clear, however, if the models include considerations of theimpact of background neutron sources to the critical states, which would bring the fluxes toconstant while the reactors were in fact at subcritical states. It was not clear either what wouldbe the result if the models do not include gaseous or volatile isotopes such as krypton or xenon,that may accumulate in different parts of the fuel rods. It is not clear either how these isotopesare treated differently for CRCs at Beginning of Cycles (BOC) and CRCs at End of Cycles(EOC). The applicant is requested to reevaluate the conclusion and demonstrate that themodels built for code benchmarking did not introduce additional bias in the criticality safetyanalyses.This information is necessary for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-19Discussions with a company that develops core design and core analysis software (Studsvik),and that has also performed many core analyses, indicate that aspects listed in the RAI, namelythe impact of background neutron sources or the role of gaseous or volatile isotopesaccumulating in different parts of rods, have a negligible effect on critical state points. Thoseaspects are in fact not modeled in the current state-of-the-art core analysis software, and thereis no indication that this results in any relevant deficiencies of those analyses. We thereforebelieve the scope of the CRC benchmarking analyses is appropriate and sufficient, specificallyin light of the limited role of the CRCs as Benchmarking experiments (see response to RAI 4-16). Note also that the input data compiled and documented by DOE, which describe andcharacterize the critical conditions, did not address those aspects discussed in the RAI. Further,we are unaware of any publicly available information that indicates these aspects could have anoticeable effect on critical state points.Regarding the bias please note that any bias that would reduce reactivity is conservativelyneglected. The fact that the CRC bias is slightly above 1.0 has therefore no impact on the safetyanalysis. However, the bias uncertainty from the CRCs is fully applied. Please see alsoresponse to RAI 4-16.NRC RAI 4-20Provide justification for modeling fully inserted control rods (CRs) with the fuel assembly havingpartially inserted CRs.Page 5 of Holtec Report, "Commercial Reactor Critical Benchmarks for Burnup Credit," states:"However, in a few cases, the model for the fully inserted CRs was used when the CRs were notPage 31 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONfully inserted, or the model for the partly inserted CRs was used when the CRs were actuallyfully inserted." The staff understands that modeling partially inserted CRs as fully insertedwould produce conservative result for criticality analysis. It is not clear, however, why a modelwith partially inserted CRs can be used for a fuel assembly having fully inserted CRs. Theapplicant is requested to provide justification for modeling fully inserted CRs with the fuelassembly having partially inserted CRs.This information is necessary for the staff to determine if the spent fuel shielded transfer caskmeets the regulatory requirements of 10 CFR 50.68, 10 CFR 72.24(c)(3), 72.24(d), and 72.124.Response to RAI 4-20The quoted section of the text from Holtec Report on the Commercial Reactor Criticals refers tomodeling details outside and above the active region of the fuel. Within the active region of thefuel, CRs are modeled accurately as stated at the end of the paragraph that contains the quotedtext, which reads "Note that the insertion position of the CRs in the active fuel region is alwaysmodeled correctly to about 1 cm". Since the modeling discrepancy occurred outside of theactive region it was considered not significant.In this context, please also see the response to RAI 4-16 on the purpose and relevance of theCRC benchmarks for the safety analyses.NRC RAI 4-21[deleted]REFERENCEReference [K.C] in the Responses to the RAls on Chapter 4 refer to the correspondingreference in the licensing report, and is also repeated here for completeness[K.C] HI-951251, Latest Revision, "Storage, Transport, and Repository Cask System (HI-STAR Cask System) Safety Analysis Report", USNRC Docket 71-9261CHAPTER 5 -THERMAL-HYDRAULIC EVALUATIONNRC RAI 5-1Verify that all thermal properties used in the analyses properly cover the expected temperaturerange during normal, loading, off-normal and accident conditions. (TCB)For example, Table 5.2.9 of the SAR includes thermal properties of steam which appear to befor superheated steam. Thermal evaluation results provided in the SAR do not indicate thepresence of superheated steam. The use of these properties to calculate the effective thermalproperties of the steam gap could overestimate heat transfer and could result in incorrectlycalculated temperatures.This information is needed for the NRC staffs review to ensure compliance with the criteriacontained in GDC 61.Page 32 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONResponse to RAI 5-1It should be noted that the thermal conductivity of superheated steam is lower than saturatedwater vapor [5-1.1]. Since superheated steam properties were used for the steam gap within theSTC, the effective thermal conductivities of the steam gap used in the licensing basiscalculations were conservative. Therefore, the predicted temperatures and pressure presentedin the licensing report were conservative.The thermal calculations presented in Chapter 5 of the licensing report have been revised andnow use the properties of materials over the full temperature range during normal, loading, off-normal and accident conditions. The steam properties have also been revised to use thethermal properties of saturated steam. The results of the modified analyses show smallincreases in the predicted peak cladding temperature for all the licensing basis analyses.In summary, the thermal properties of all materials in the licensing report have been revised tocover the expected temperature range.Reference[5-1.1] "Fundamentals of Heat and Mass Transfer", Frank P. Incropera and David P.DeWitt,fourth edition.NRC RAI 5-2Justify the emissivity value of the water surface used to calculate the effective thermalconductivity of the steam gap region inside the STC. Provide also the reference where thewater emissivity is obtained. (TCB)The staff needs to verify realistic parameters are used to properly characterize regionsrepresented by effective properties which would assure a realistic or conservativerepresentation of the heat transfer characteristics of the system. Table 5.2.1 provides thethermo-physical property references for all materials used. Note 1 on this Table states thewater emissivity is not reported, as radiation heat dissipation from these surfaces isconservatively neglected. However, water emissivity is used to calculate effective thermalconductivity of air and steam spaces.This information is needed for the NRC staff's review to ensure compliance with the criteriacontained in GDC 61.Response to RAI 5-2The emissivity of water was used in the analyses but was not reported in the licensing report.The emissivity of water used in the analysis was 0.96 [5-2.1] and this value has been added toTable 5.2.4 of the licensing report. It should be noted that the emissivity of water used todetermine the radiation heat transfer [RTI. Note that thefootnote in Table 5.2.1 has been modified to clarify that water is opaque to thermal radiation.

Reference:

[5-2.11 D. Q. Kern, "Process Heat Transfer", McGraw Hill Kogakusha, (1950).NRC RAI 5-3Page 33 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONClarify why the SAR reports lower temperatures compared to the thermal evaluation included inthe original application. (TCB)The staff needs to verify how the updated thermal evaluation resulted in lower temperatures(e.g., if model conservatisms were removed, provide adequate justification, etc.)This information is needed for the NRC staffs review to ensure compliance with the criteriacontained in GDC 61.Response to RAI 5-3The temperatures reported in the licensing report (Revision 3) are lower than those reported inthe original application (Revision 1) due to the following reasons:2. Air below the STC lid was replaced with steam. The emissivities of surfaces inside theSTC remained unchanged. The thermal conductivity of steam is lower than that of air.However, the effective thermal conductivity of space below the STC lid was lower by asmall amount (less than 1%) when this space is occupied by steam instead of air. Thischange had no significant impact on temperatures.3. The net heat balance of the system in the original application was higher thanHDJ which lead to higher temperatures.In summary, the changes mentioned above caused the reported drops in the predictedtemperatures. Please note that the temperatures and pressures have changed again due to therevised thermal analyses reported in the licensing report. The reasons for these changes areprovided in the responses to RAIs 5-1, 5-4 and 5-10.NRC RAI 5-4Revise all thermal analyses to assure energy balance has sufficiently converged to provideassurance calculated temperatures are representative of all applied thermal loads. (TCB)When performing analysis audits of some of the calculations, the staff noticed there is a heatimbalance of about 7% which indicates the calculation is not fully converged which could resultin lower non-conservative temperatures.This information is needed for the NRC staffs review to ensure compliance with the criteriacontained in GDC 61.Page 34 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONResponse to RAI 5-4The thermal analyses have been revised to assure that the energy balance has sufficientlyconverged. To ensure that steady state conditions were achieved in these analyses asufficiently high number of iterations was utilized and the peak temperatures of the fuel and STCcomponents were tracked until it was concluded that the peak temperatures of components didnot change significantly with any further iterations.To achieve a greater heat balance the following changes were made in the analyses:1. [PROPRIETARY TEXT REMOVED].2. D_______EXT R E49,.RLP T I I NEEQhese analyses also incorporate the changes made tothe material properties in accordance with the response to RAI 5-1. The peak claddingtemperatures along with the heat balances are reported below:No. of cells in No. of cells in Heat Balance PeakDescription radial direction of the radial the axial on outer Claddingthe Annulus direction of direction of surfaces of Temperaturethe A Water Jacket Fuel the HI-TRAC (°C)Mesh 1Notel kL,Mesh 2Mesh 3Note 1: This mesh was used in the licensing basis calculations with design basis heat in revision 3 of theLlicensing report. ýkR-Yport.]1The results show that the mesh used for the calculations in Revision 3 of the licensing reportpredicted higher temperatures. The peak temperatures due to mesh refinement and with abetter heat balance are lower than the peak temperatures obtained for the mesh used inRevision 3 of the licensing report. All the steady state cases presented in this licensing report,therefore, continue to use Mesh 1. For cases that use IPROP Y1 TE-iI_M_6Efmodel (e.g. simultaneous loss of water in the HI-TRAC annulus and the waterjacket), iRRTTE. Iffil For all the transient analyses, Mesh 3 is usedsince it results in a mesh independent solution.

Reference:

[5-4.1] Section 25.4.3 of FLUENT 6.3 User's Guide.NRC RAI 5-5Perform the transient pressure rise for at least 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to make sure the pressure increaseused to monitor for fuel misload is approaching steady state. (TCB)Page 35 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONFigure 5.3.2 of the SAR indicates a pressure increase of about 4.2 psi is expected for the first24 hours. The staff needs to have additional assurance the pressure increase is converging tothe steady state value which would indicate a safe onsite transfer may proceed.This information is needed for the NRC staff's review to ensure compliance with the criteriacontained in GDC 61.Response to RAI 5-5A transient STC pressure rise calculation has been performed for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months assuming a designbasis heat load. A mesh independent solution has been achieved by performing meshsensitivity studies as discussed in the response to RAI 5-4. Since Mesh 3 provides a meshindependent solution, the transient pressure rise has been determined using this mesh. Thepressure rise curve is shown in Figure 5.3.2 of the licensing report and in Figure 1 below. TheSTC loaded with fuel assemblies and filled with water has a high thermal inertia.*PqRPf:M AI .1. It is observed that the rate of increase of thetemperature inside the STC and the STC pressure with time is extremely small over the 24-hourduration analyzed, of the order of [P, 7 Iffi Therefore, thisprovides assurance that the solution will not be altered by the use of a smaller time step. Theabove mentioned features of the thermal model ensure that the uncertainties in the transientanalysis are small and will not alter the solution as presented.Figure 1: 48-hour STC Pressure Rise under Design Basis Heat LoadIt should be noted that as a result of the reanalyzes performed in response to RAI 5-4 the 24hour expected STC pressure rise has decreased fro TEXT RhMOVEDI.The pressure increase at the end of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is [RR-IP TARY TEXT REMOVED]?. Therate of pressure rise at the end of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months indicates that a steady state condition has not beenreached yet; however, the rate of pressure rise over the 48-hour duration is only approximately[PROPT 'MVD. If it is very conservatively assumed that this rate ofpressure rise continues for an additional 28 days (30 day VCT breakdown accident condition)then the total STC pressure rise would be limited to approximately 'PRQPRIEfARY TF'TREMOVED]. As the initial STC pressure would be sub atmospheric the maximum expectedSTC pressure would be [kOP: [A jr]txjL EDý which is lower than the STCdesign pressure limit of 50 psig for normal operations and significantly below the applicableaccident limit of 90 psig. As noted this is a very conservative evaluation as it neglects theincrease in heat loss from the system with increasing STC water temperature. A more realisticanalysis shows, under design basis assumptions as reported in Table 5.3.2 of the licensingreport, that the steady state STC pressure would be LPROPRJETARyhMQýYJO jdemonstrating significant additional margin to the STC design limits.In summary, even though the STC pressure increase has not converged to the steady statevalue the rate of pressure at 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is small indicating that a safe onsite transfer mayproceed.This information has been added to Section 5.3.4 of the licensing report.Page 36 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONNRC RAI 5-6Perform a transient pressure rise calculation for the case when a possible misload has occurredbased on heat loads representative of the IP3 spent fuel fuel inventory to assure the pressurerise can be managed by the licensee's in-place operating control procedures. (TCB)The transient pressure rise provided in the SAR applies only for design basis heat load. Thestaff needs to have assurance the licensee has the capability to implement corrective actions, ifneeded, in the case of an occurrence of a misload.This information is needed for the NRC staff's review to ensure compliance with the criteriacontained in GDC 61.Response to RAI 5-6[PJW c Y 1Ifi .The heat load for a fuel assembly that has beensubcritical for at least 90 days with a bounding burnup of 55,000 MWD/MTU, was determined tobe E [5-6.1]. Proposed TS 4.1.5.4 dictates that transferoperations shall only be conducted when the irradiated fuel assemblies in the Unit 3 spent fuelpit have been subcritical for at least 90 days.The hypothetical accident condition assumes that a single fuel assembly with a heat load ofyWV ...... is mis-loaded in the STC basket. Mesh 3 has been usedin this analyses (as discussed in the response to RAI 5-4) since it provides a mesh independentsolution. The misloaded fuel assembly was conservatively placed in the inner cell region tomaximize STC temperatures and, consequently, pressure rise inside the STC. All other STCbasket storage locations were assumed to be at their design basis heat load. PR IEREL E D. The transient pressure rise inside the STC for the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months afterloading, both for the design basis heat load and the severe mis-load accident conditions, arepresented in Figure 1 below and included as Figure 5.3.3 in the licensing report. Likewise, therate of change of STC pressures are presented in Figure 2 below and included as Figure 5.3.4in the licensing report. Figure 2 shows that a rate of change of tPP,rRORTARY TE~iRgEMOVEDcan be used to differentiate between a design basis heat load and a severemisload shortly after commencement of the STC pressure rise surveillance.The response to RAI 1-1 provides justification that the [PPY§EI:i1RýEIQyEJPDSTC pressure rise acceptance criterion, over a rolling 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period, ensures that asevere fuel misload would be detected by the pressure monitoring system within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . If asevere fuel misload is detected, corrective actions that could include unloading fuel from theSTC to the spent fuel pool would be performed. These actions are described in Subsection10.2.3 of Chapter 10 of the licensing report and are included in the TS. These actions provideassurance that the capability exists to implement corrective actions, if needed, in the case of anoccurrence of a severe fuel misload. This information has been added to Section 5.3.4 of thelicensing report.

Reference:

[5-6.1] "Spent Nuclear Fuel Source Terms", Holtec Report HI-2022847, Revision 5 (HoltecProprietary).Page 37 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONFigure 1: Comparison of STC Pressure Rise between Design Basis Heat Load and Severe FuelMisload AccidentFigure 2: Comparison of Rate of Change of STC Pressure with Time between Design BasisHeat Load and Severe Fuel Misload AccidentNRC RAI 5-7Review the thermal properties used to determine the thermal conductivity of saturated watervapor as a function of temperature. (TCB)Page B-7 of Holtec report HI-2084146 provides saturated vapor thermal conductivity as afunction of temperature. Thermal conductivity values tabulated at 371 C and 372 C appear tocorrespond to saturated liquid water. This may increase the effective conductivity at thesetemperatures.This information is needed for the NRC staffs review to ensure compliance with the criteriacontained in GDC 61.Response to RAI 5-7The thermal properties used to determine the thermal conductivity of saturated water vaporhave been reviewed and it has been confirmed that the thermal conductivity values on page B-7of Holtec report HI-2084146 are for saturated vapor. Table A.11 of the listed reference providesseparate values for saturated liquid and saturated vapor, both as functions of temperature. Asthe temperature approaches the supercritical region (beginning at 3710C) the saturated vaporconductivity rises to approach that of saturated liquid. This is readily apparent in the followingplot of the data provided by the reference. There is typically little difference in thermalconductivity between saturated liquid water and saturated water vapor in the supercriticalregion.Page 38 of 94 800700600500E 400E3002001000HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONThermal Conductivity of Liquid Water and Water Vapor(from Table A.II of lAPS Formulation 1985 for the Thermal Conductivity of Ordinary Water Substance)~Start of SupercriticalRegion at 371 C100150200250Degrees C300350400-Liquid -Vapor

Reference:

[5-7.1] "IAPS Formulation 1985 for the Thermal Conductivity of Ordinary Water Substance."NRC RAI 5-8Clarify how the STC pressure rise would be controlled if, based on the thermal analysis for thisconfiguration, it is predicted that water would be boiling under normal conditions of transfer.(TCB)Based on auditing of some of the thermal calculations (see also RAI 5-4 above), the staffnoticed there is not adequate convergence in the heat balance. For a properly convergedsolution (maximum temperatures reached, adequate heat balance), the licensee could predicthigher temperatures with water reaching the boiling point.This information is needed for the NRC staff's review to ensure compliance with the criteriacontained in GDC 61.Response to RAI 5-8The thermal analyses have been re-performed as documented in the revised licensing report toaddress the RAIs above and have achieved greater numerical convergence and a better heatbalance. The results of these analyses for the normal conditions of transfer show that bulkboiling of the water within the STC will not occur.The water and vapor (no air) inside the STC form a closed self-equilibrium system. The waterPage 39 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONsurface temperature is always equal to the local saturation temperature at the water surface.The local water saturation temperature below the water surface increases with water depth dueto an increase in the local pressure.In a small region in the center of the STC, that is less than 3 feet deep, the water temperature isslightly (-20C) higher than the local saturation temperature. The phenomenon of boiling mayoccur in this small region. However, it is to be noted that the fuel in this region is inactive part ofthe fuel assembly. The bubbles due to local boiling will rise upwards and condense due to lowertemperatures near the water surface. This may result in a slight increase in the surfacetemperature of water inside the STC, thereby resulting in a slight increase in the STC pressure.Then the local boiling will be suppressed due to pressure increase inside the STC. Furthermore,the enhancement of heat transfer due to phase change will decrease PCT and volume of waterwhich exceed the local saturation temperature. Since this is a closed system, the watertemperature and vapor pressure will reach an equilibrium state.It should be noted that the phase change is not modeled in the thermal analyses and thereforethe actual water surface temperature may be a little higher than the computational result, but itcannot go beyond the PCT inside the STC. Consequently, the maximum possible STC pressureis the saturation pressure corresponding to peak computational temperature inside the STC.The saturation pressure corresponding to the peak temperature within the STC during designbasis normal transfer conditions was determined to be A-- ETA7ý1 E 1X R EM A,which is still significantly lower than the design STC pressure of 50 psig specified in Chapter 3of the licensing report.NRC RAI 5-9Describe the code assessment conducted for the software models used to calculate the thermalresponse of the STC and HI-TRAC. (SBPB)Section 5.3, "Thermal Evaluation of Fuel Transfer Operation," of the SAR included a descriptionof the three-dimensional modeling used for thermal evaluations. The discussion described thegeneral modeling and some conservative assumptions included in the model. In addition, theresponse to NRC staff requests for additional information provided in Attachment 1 to the letterdated October 5, 2010, indicated that the models were used for similar evaluations of spent fuelstored in spent fuel pools. However, in this application the model was used to evaluateconditions involving natural convection heat transfer in air environments and to evaluateradiation heat transfer. Section 15.0.2, "Review of Transient and Accident Analysis Methods,"of the NRC Standard Review Plan (NUREG-0800) describes specific areas of review for modelsused for transient analyses. The review areas include code assessment, which the staffdescribed as a complete assessment of all code models against applicable experimental dataand/or exact solutions in order to demonstrate that the code is adequate for analyzing thechosen scenario.Provide a code assessment for the accident scenarios involving natural convection heat transferin air environments and radiation heat transfer that demonstrate the adequacy of the model forthose scenarios.This information is needed for the NRC staff's review to ensure compliance with the criteriacontained in GDC 61.Page 40 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONResponse to RAI 5-9FLUENT is the principal code deployed to evaluate scenarios involving natural convection heattransfer in air environments and radiation heat transfer. This code was benchmarked in the late90s to evaluate similar scenarios in support of Holtec's dry storage applications licensed under10CFR72 (on-site storage) and 10CFR71 (transportation). FLUENT was also deployed toevaluate scenarios involving natural circulation of water in spent fuel pools to demonstratethermal-hydraulic requirements for safe storage of spent nuclear fuel in water environment. Apartial list of Dockets under which the code has been used for safety analysis of casks is asfollows:Application Docket No.HI-STAR 100 Storage 72-1008HI-STAR 100 Transportation 71-9261HI-STAR 60 Transportation 71-9336HI-STAR 180 Transportation 71-9325HI-STORM 100 Storage 72-1014HI-STORM FW 72-1032St. Lucie Unit 1 50-425Diablo Canyon Units 1 and 2 50-275, 50-455Beaver Valley Unit One 50-334The FLUENT code is validated using data from PNNL tests conducted with the TN-24Pprototype cask loaded with irradiated SNF. The FLUENT code is benchmarked withindependent COBRA-SFS thermal analysis of the HI-STORM cask by PNNL. The codevalidation and benchmark work is archived in the QA validated Holtec report, "Topical report onthe HI-STAR/HI-STORM Thermal Model and its Benchmarking with Full-Scale Cask Test Data",HI-992252, Rev. 1. A succinct description of the code validation work and results confirming theadequacy of the FLUENT computer code is provided below.The TN-24P test cask is a 24-cell prototypical metal cask designed to store PWR fuel. The caskthermal tests were conducted under an EPRI sponsored work' by Pacific Northwest Lab (PNNL)and Virginia Power Company (VPC) at the Idaho National Engineering Lab (INEL). The testcask was loaded with irradiated PWR fuel from the Surry reactor. The cask was tested withsignificant decay heat (20.6 kW) under an array of scenarios wherein one or more of theprincipal modes of heat transfer -conduction, natural convection and radiation -were active.The test scenarios are listed below.Case no. Scenario Principal Heat Transfer Modes1. Cask vertical, Vacuum Radiation2. Cask horizontal, Vacuum Radiation, Conduction3. Cask horizontal, Nitrogen filled Radiation, Conduction4. Cask vertical, Helium filled Radiation, Conduction5. Cask vertical, Nitrogen pressurized Natural Convection, Radiation6. Cask vertical, Helium pressurized Natural Convection, RadiationThe principal results obtained from the validation study are summarized below:Note 1 "The TN-24P PWR Spent Fuel Storage Cask: Testing and Analyses", EPRI NP-5128, April 1987.Page 41 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONtjl- TAR1FwPThe above results support the conclusion that FLUENT code provides conservative assessmentof cask thermal condition and confirms its suitability for safety analysis of casks loaded withirradiated SNF.The COBRA-SFS analysis of the HI-STORM storage cask by PNNL is included as Attachment 1of the Holtec benchmarking report discussed above. The principal results of the FLUENTbenchmarking study are graphed in the Figure below. The Figure provides additional assuranceof the adequacy of the FLUENT code to yield conservative assessment of cask thermalcondition.Comparison of FLUENT Peak Clad Temperature Solution with PNNL ResultNRC RAI 5-10Discuss the effect of the centering assembly on heat transfer from the STC to the HI-TRAC andthen to the atmosphere during scenarios involving loss of water from the HI-TRAC annulus.(SBPB)Holtec Report No. HI-2084146, "Thermal Hydraulic Analysis of IP3 Shielded Transfer Canister,"provided information about analyses of various scenarios evaluated for passive rejection ofdecay heat to the environment. Section 7.4 discussed the analysis of the simultaneous loss ofwater from the water jacket and HI-TRAC annulus, and mentioned the use of the DiscreteOrdinates radiation heat transfer model in FLUENT. This analysis credited radiation and naturalconvection heat transfer within the annular space between the STC and the HI-TRAC.The annular space between the STC and the HI-TRAC contains the HI-TRAC/STC centeringassembly. The staff believes the centering assembly may adversely affect heat transfer duringthe scenario involving loss of water from the HI-TRAC annulus because it would interfere withinternal radiation heat transfer and reduce internal natural circulation air flow. Describe in detailthe internal natural circulation in air and radiation heat transfer models for the loss of water fromthe HI-TRAC annulus scenario. Specifically address how the effect of the centering assemblywas incorporated in the models.This information is needed for the NRC staffs review to ensure compliance with the criteriacontained in GDC 61.Response to RAI 5-10A thermal model that explicitly includes the HI-TRAC/STC centering assembly (12 total) hasbeen developed to evaluate its effect on the temperature and pressure of the system. Thethermal analysis of the scenario involving the loss of water from the HI-TRAC annulus and waterjacket has been performed with the centering assembly explicitly included in the model. Anemissivity of 0.2 [5.10-1] was used for the Aluminum centering assembly surfaces.W.. A steady state analysis was performed and thecomponent temperatures, STC and HI-TRAC internal pressures are reported in Tables 5.4.5Page 42 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONand 5.4.6 of the licensing report. The results show [PROP APY TEX17EMOTED inpeak cladding temperature due to the presence of centering assemblies. The STC internalpressure [PROqET T T '_""'. The effect of including the centering assemblyon temperatures and pressure is small compared to their respective safety margins to theaccident limits, which are provided in Tables 3.1.1 and 3.2.1 respectively.The details of the analysis and its results are summarized in the table below and have beenincluded in Section 5.4.3 of the licensing report.

Reference:

[5-10.1] "Fundamentals of Heat and Mass Transfer", 4th Edition, F.P. Incropera and D.P. DeWitt,John Wiley & Sons, Inc., New York, 1996.SIMULTANEOUS HI-TRAC ANNULUS AND JACKET WATERLOSS ACCIDENT TEMPERATURESCHAPTER 6 -STRUCTURAL EVALUATION OF NORMAL AND ACCIDENT CONDITIONLOADINGS6-1. Provide the technical basis for ignoring the water inside the hollow aluminum tubes whenmodeling the tube assembly for mitigating side impact effects associated with the non-mechanistic tipover of loaded HI-TRAC cask analysis. (SMMB)SAR Section 6.2.8, Rev. 3 states, "the tip-over of the transfer cask has been carried out withoutconsidering the fluid coupling (cushioning) effect of water inside the annular space around theSTC and within the centering assembly tubes." It is not entirely clear why water inside of thehollow aluminum tubes is not included in the model, given that the water incompressibility mayadd significant stiffness to the tubes during a tipover event.This information is required to demonstrate that the system can withstand the worst-case loadsand successfully preclude an unacceptable release of radioactive materials to the environment,in compliance with GDC 61 and 10 CFR 72.122.Response to RAI 6-1:The reviewer is correct in the observation that neglecting the effect of water inside the impactlimiter tubes may underestimate the crush resistance of the tubes. It is indeed true that a closedend tube full of water will be considerably stiffened by the captive water because water isessentially incompressible. For this reason, the ends of the impact limiter tubes in the HI-TRACsystem are designed to be fully open and un-constricted so as to permit free flow of water out ofthe tube from both ends if the tube were to be squeezed by a lateral force. Nevertheless, underthe postulated tip-over scenario (Load case 9 in Section 6.2.8 of the Licensing Report) a part ofthe kinetic energy of the STC will be expended in hydrodynamic squeezing of the water insidethe impact limiter tubes during their crushing under the inertial momentum of the STC'sinternals. The hydrodynamic squeezing of water translates to a net increase in velocity head ofthe water. It is intuitively obvious that the velocity of the water stream escaping from the tubePage 43 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONends will increase as the opening begins to ovalize under the crush load. Thus the kineticenergy associated with the escaping water will increase as the crushing of the tube endsproceeds. This phenomenon continues until the crushing process stops. If the ends of the tubewere to close completely then the apparent stiffness of the tube against crushing willundoubtedly increase markedly. However, as can be seen from Figure A (extracted from theLS-DYNA run files), the extent of crushing is localized near the top end of the impacted tubes(only one tube is subject to any visible crushing) and the extent of crushing, indicated by thereduction in the diameter is less than 4% (conservatively) of the overall tube length. An estimateof the hydrodynamic consequence of this local deformation can be computed using theprocedure outlined in Levy & Wilkinson2 which suggests the use of the principles of classicalhydrodynamics to estimate the associated expenditure of mechanical energy.Following Levy, et al, a first order estimate of this effect can be obtained by the simpleexpedient of computing the energy extracted from the system during the crush event by theexpulsion of the water as the STC advances towards the HI-TRAC surface.The total volume of water expelled from the two ends of the aluminum tube, Ve, is obtained bysubtracting the crushed inner space of the tube from the uncrushed (virgin) volume. The velocityof ejection of water from the tubes, v, is given byTime of crush (Figure B) u = 0.043 secInside area of tube cross-section A = rr (6.5 in)2/4 = 33.18 in2Height of water in the tubes in vertical config. L = 179.75 inTotal volume of water inside a single tube V = L x A = 5964 in3Volume of water expelled from the tube Ve = 238.6 in3(approx. 4% of length "U, Figure A)Density of water (approx.) p = 62.4 lbf/ ft3 = 0.036 lbf/ in3Average open area at top end Aav = A/2 = 16.59 in2Velocity of the water v = Ve/(Aav T) = 334.5 in/sec.where T is the duration of crush, and Aav is the time-averaged open area at the top end of thetube through which water is squeezed out. The calculated value for Aav conservatively assumesthat the top end of the tube is completely crushed (A = 0) at time 'r. Furthermore, the abovevelocity calculation takes no credit for any exit flow of water through the bottom end of the tubeeven though it is substantially open.The total energy E carried away by the squeezing of water is given by,E P Ve3 = 1243 Ibf-in2 Aav2 r2g,It is seen from the above, the energy extracted by the expulsion of the water inside the tube isminiscule (less than 0.06%) compared to the total kinetic energy of 2.23e6 lbf-in (see Figure C)of the STC at the onset of the impact, which must be dissipated during the impact event.Thus it is concluded that the stiffening effect caused by the energy required to expel the2 See page 344 "Component Element in Dynamics by Levy & Wilkinson, McGraw Hill, 1976".Page 44 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONcontained water from the tubes to satisfy (hydraulic) continuity is extremely small (a second orhigher order effect) for the specific case of the HI-TRAC tip-over event.Based on the above, it can be concluded that the effect of neglecting the stiffening effect ofwater in the tubes is inconsequential and is more than offset by other explicit conservatisms inthe model such as the assumption of a rigid STC shell and doubling of the HI-TRAC's radialplate thickness subject to direct impact, among others, as discussed in the Licensing Report,Table 6.2.8.2.IPROPRIETARY FIGURESRE VDCHAPTER 7 -SHIELDING DESIGN AND ALARA CONSIDERATIONSThe proposed amendment seeks to perform a wet transfer of spent fuel from the IP3 SFP to theIP2 SFP using an STC. The STC functions as a transfer cask in these operations and may beconsidered a lightweight transfer cask since it is designed for a limited-capacity crane thatcannot handle an approved transfer cask like those used for approved spent fuel dry storagesystems. The STC presents some unique shielding and radiation protection considerations,with significantly higher dose rates for the proposed allowable contents. The STC must be usedin conjunction with a HI-TRAC transfer cask, which provides additional shielding so thattransfers between FSBs as well as preparation for the transfer and unloading of the STC can bedone in a safe manner. This is a new, first-of-a-kind review for a system and operations of thiskind.Normally, transfer casks provide sufficient biological radiation shielding such that workers maysafely be in the vicinity of the transfer cask. This does not appear to be the case with the STCdesign. The staff is not only concerned about occupational doses during normal, off-normal,and accident conditions, but also public doses. Since this is an amendment under 10 CFR Part50 and the action is limited to the Indian Point Energy Center site, considerations are given withregard to the site features. Still, the staff must ensure that enough controls are in place toprovide reasonable assurance that both the public and occupational dose limits in 10 CFR Parts20, 50 and 100 (via compliance with the intent of Part 72 limits) are not exceeded.The following RAIs are geared towards obtaining enough information so that the staff may makea determination regarding whether there is reasonable assurance that the STC may be usedsafely and in accordance with the regulations. The staff is particularly concerned about apotential off-normal event involving either the hang-up of the crane or a malfunction of theremote handling equipment (if used). Crane hang-ups are not uncommon, especially whencranes are loaded with weights approaching their capacity limits. Additionally, the staff is askingfor clarification and/or additional information to ensure that the license, TS, and licensing reporteach contain the appropriate level of information needed to control the design basis for thisunique transfer system.General Response to Shielding Design and ALARA RAIsEntergy recognizes the high importance of the shielding design and ALARA considerations forthe STC, and the concerns about doses and dose rates from the system. In response to thosedose concerns for the STC, the previous calculations and results have been reviewed. ThePage 45 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONconclusion of the review is that some of the high dose rates that were previously reported werethe result of overarching conservatisms, and are not indicative of the doses to be expected fromthe system. An effective radiation protection approach for the STC can only be based onmeaningful calculated doses and dose rates, such that dose rates are conservative but at thesame time realistic. Accordingly the analyses have been revised to be more consistent with thisapproach, A further enhancement is that the sealing system of the STC will be tested to ANSI14.5 leak tight criteria, which eliminates concerns regarding effluent doses from the STC.Following is a brief summary of revisions to the dose analyses, while the individual RAIresponses below and the revised licensing report provide additional detail. Note that for someaspects of the analyses, a more realistic approach is taken for the doses and dose rates fromthe bare STC, while more conservative assumptions are made for dose rates from the STClocated inside the HI-TRAC. The more conservative assumptions for the HI-TRAC simplify theanalyses, and can be used since dose rates from the HI-TRAC are low:Design Basis Assemblies: The STC is designed for regionalized loading in both thethermal and shielding designs. However, while the previously proposed TS clearlydefined the heat load limits for each region, only a single upper bound burnup and asingle lower bound cooling time were defined.. The combination of those values in theshielding analysis resulted in dose rates that were up to a factor 5 higher than those forany loading configuration that could actually meet the thermal requirements. To addressthis, the revised proposed TS now contain explicit burnup, cooling time and enrichmentlimits for the two regions of the basket, and those limits are the sole basis for the doseevaluations. Assembly heat load, bumup and cooling time are selected in a consistentfashion for each of the two regions, and based on the inventory of fuel assemblies to betransferred.* Water in the STC: The previously reported dose analyses assumed unborated waterinside the STC. However, for the STC, a minimum soluble boron level of 2000 ppm isspecified in the proposed TS. This boron level substantially reduces the neutron dosefrom the cask, and is now used in all STC dose analyses3, when the STC is outside ofthe HI-TRAC. Unborated water is used for the HI-TRAC dose evaluations." Azimuthal dose variations: Azimuthal dose variations were evaluated for the STC, butonly the peak dose rates were reported in the licensing report. Minimum, maximum andaverage dose rates are now reported, together with axial and radial variations indicatingin which areas minimum and maximum dose rates are to be expected from the bareSTC. However, maximum dose rates considering the azimuthal variance are reported forHI-TRAC.* Material Thicknesses: Since the STC has been manufactured, as-built materialthicknesses are used in the dose analyses for the bare STC, while minimum materialthicknesses are used for the analyses of the STC inside the HI-TRAC.3 Note that no changes were made to the criticality calculations in that respect, i.e. soluble boron in criticalitycalculations is only credited under accident conditions, but not under normal conditions.Page 46 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONEffluent Dose: The STC seal system will be tested as leak-tight per ANSI 14.5.Consequently, effluent is no longer expected, and effluent dose rates are no longerconsidered in the analyses.In summary, more realistic dose rates are evaluated for the bare STC, while more conservativeapproaches are taken for the HI-TRAC dose evaluations. Additionally, the maximum doselocations for the bare STC are clearly identified. In general, STC dose rates are a factor of 3 to 5less than the dose rates reported for the bounding loading combination of the previoussubmittal. Dose rates for the bare STC are therefore comparable to those for standard transfercasks used in the industry. For example, the average dose rate on the radial surface of the STCat mid height is about 3 rem/hr, while the corresponding value for the HI-TRAC 100 in HI-STORM 100 FSAR is about 4 rem/hr, both with design basis fuel. However, actual fieldmeasurements are expected to be much lower than the estimated dose rates based on theexperience with the HI-TRAC. Additionally, in response to the shielding design RAIs Entergy isproposing controls that will provide reasonable assurance that both the public and occupationaldose limits in 10 CFR Parts 20, 50 and 100 (via compliance with the intent of Part 72 limits) arenot exceeded. Compliance with the intent of Part 72 is demonstrated by adopting the Part 72numerical limits where appropriate.NRC RAI 7-1Justify the homogenization of the fuel assembly with the moderator in the shielding model.(CSDAB)Question 7-16 in the first round RAI dealt with accounting for neutron multiplication and its effecton dose rates with an appropriate geometry. Modeling assumptions regarding fissile materialand moderator can influence the neutron multiplication of a system, with homogenized modelsunder-predicting multiplication occurring in heterogeneous systems of fissile material andmoderator. The applicant's response does not address this question and instead assumes a k-effective value to define multiplication without consideration of how k-effective differs between aheterogeneous and a homogeneous system of the same materials or how the different systemsbehave from a shielding perspective. Further, the response only considers the loaded HI-TRAC; it should also include the impact on dose rates for the loaded STC outside of the HI-TRAC.This information is needed to confirm compliance with 10 CFR 20.1101(b), 10 CFR50.90 and50.34a(c) and the intent of 10 CFR 72.104, 72.106(b) and 72.126(a).Response to RAI 7-1To further justify the homogenization of the fuel assembly with the moderator in the shieldingmodel, additional studies have been performed. The cases in the study, including the designbasis case, are described below. For each case, the specific modeling details that are importantto the neutron multiplication effect are specified. Note that all other aspects of those calculationsare identical between the cases, i.e. are identical to the design basis calculations.A -Design Basis Case-Homogenized Fuel Model-Fuel assembly assumed fresh, 3.6 wt% EnrichmentPage 47 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONB -Heterogenuous Fuel Model-Fuel assembly assumed fresh, 3.6 wt% EnrichmentC -Heterogenuous Fuel Model-Fuel assembly modeled as spent fuel, 45 GWd/mtU, 5 years cooling timeThe average STC and HI-TRAC surface and 1 m dose rates are as follows:Homogeneous Heterogeneous HeterogeneousFuel Model, Fuel Model, Fuel Model,Location Fresh Fuel Spent Fuel Fresh Fuel(mrem/hr) (mrem/hr) (mrem/hr)(Case A) (Case C) (Case B)STCSurface 586.4 350.7 586.81 m 142.7 87.5 144.0HI-TRACSurface 0.97 0.32 Not evaluated1 m 0.38 0.15 Not evaluatedNote that the total dose rates reported in the above table only comprise of neutron and capturedgamma, since dose contributions from other sources (fuel gamma, Co-60 gamma) areessentially unaffected by the fuel modeling approach.The results show that a heterogeneous fuel model results only in a marginally higher dose ratewhen the fresh fuel assumption is maintained. Typically, higher multiplication factors, and thushigher neutron dose rates, are expected from heterogeneous models. However, this isapparently offset by the presence of the soluble boron credited in the models. Accounting forfuel burnup instead of assuming fresh fuel results in a significant reduction in the neutron doserate, although the relative effect on the total dose is much less since the neutron dose is not thedominant dose contributor. Overall, this shows that using a homogenized fuel model with freshfuel is appropriate and conservative. This study is reported in Reference [L.G].NRC RAI 7-2Provide a dose rate evaluation for the STC with all shielding materials at their minimumthickness specified in the proposed TS, Appendix C, Part I, Section 1.0, addressing the impactson the occupational and public doses. (CSDAB)Based upon the sample input file provided as part of the RSI response, the current dose ratecalculations are based upon nominal steel dimensions and minimum lead thickness. While notas strong as lead, steel is still a significant shield material. For example, the half-valuethickness of steel for 1.0 MeV photons is about 1.5 cm and for 1.5 MeV photons is about 1.8cm; thus, the difference in steel between a minimum thickness present and a nominal thicknessPage 48 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONpresent is nearly 1 half-value thickness at this gamma energy; indicating the potential forsignificant differences in dose rates. Given the high dose rates for the STC with therepresentative loading, the impact on dose rates could be significant and lead to higherestimated occupational and public doses and/or the need to further modify operations to reducedoses and keep them as low as reasonably achievable (ALARA).This information is needed to confirm compliance with 10 CFR 20.1101(b), 10 CFR 20.1301(a)and (b), and the intent of 10 CFR 72.104 and 72.106.Response to RAI 7-2With respect to manufacturing tolerances of the STC, specifically the tolerances of the radialsteel, it is important to note that larger than usual tolerances were selected to provide theflexibility that is necessary to ensure the crucial weight limit of the system is met. As the STChas now been already manufactured in the shop, as-built dimensions are used for doseevaluations when the STC is outside of the HI-TRAC. The only difference found between the as-built and the nominal dimension is the thickness of the STC inner shell. Nominal dimension ofthe STC inner shell is 1", whereas the as-built dimension is 3/4" with 3/16" of weld overlay. Asmachining is performed on this weld overlay, the 3/16" thickness of the weld overlay isneglected. The revised licensing report Tables 7.4.1 to 7.4.8, which report dose rates from thebare STC, assume an STC inner shell thickness of 3/4". However, minimum STC dimensionsare applied for all the evaluations when the STC is within the HI-TRAC. HI-TRAC dose ratesare reported, with minimum STC dimensions, in Tables 7.4.9 to 7.4.22.The proposed TS have been revised to include the as built thickness of the STC inner shell. Inaddition to the manufacturing records for the individual parts of the STC, a simple verificationhas been performed based on the weight of the as-built STC (empty, without lid) in comparisonwith the weights that would be consistent with nominal and minimum radial thicknesses. Theresults of this comparison are as follows:Condition Weight (Ibs)Nominal Dimensions 44,200Minimum Dimensions 39,500As-modeled (Nominal 42,200Dimensions, except 3/4" for innershell)As-built (measured) 44,400The comparison shows that the as-built weight exceeds the as-modeled weight, and is in factclose to the nominal dimensions. The as-built weight is also significantly higher than the weightfor the minimum dimensions. Using the as-built dimensions in the model and neglecting theweld overlay as discussed above is therefore appropriate and conservative.NRC RAI 7-3Page 49 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONJustify the changes to the dose rates reported in the tables in Section 7.4 of the STC licensingreport, correcting the reported dose rates and dose evaluations as necessary. (CSDAB)In response to staffs RAIs, the applicant modified the dose rates reported in the licensing reportfor the STC and HI-TRAC. However, the changes appear to be inconsistent. For example, thesurface dose rates on the STC for the representative loading case are nearly double theirpreviously reported values; however, a number of the dose rates reported at the axial surfacesand at distance from the axial and radial surfaces have either negligibly changed or havedecreased compared to the previously reported values. Correct dose rate values should beprovided, and changes in dose rates should be adequately explained and justified. The doseevaluations in the report should also be updated, as necessary, to account for the correct doserates.This information is needed to confirm compliance with 10 CFR 20.1101(b), 10 CFR 20.1301 (a)and (b), and the intent of 10 CFR 72.104, 72.106, and 72.126(a).Response to RAI 7-3In the previous submittal, in response to RAI 7-5, azimuthal tallies were added and only theazimuthal peak dose rates were reported for the surface of the STC and at 0.5 m from the STC.This resulted in the observed increase in dose rates on and near the surface of the STC. For allother distances average dose rates were provided which resulted in a negligible or slightchange in those dose rates.In the current submittal, Chapter 7 of the licensing report has been substantially revised and thetables contain new dose rates that reflect the current analysis and are both accurate andconservative. The new submittal. now clearly identifies the average dose rates, azimuthal doserate distributions and peak dose rates, and axial dose rate distributions and peak dose rates.NRC RAI 7-4Provide dose evaluations for crane hang-up and other off-normal conditions occurring with aloaded STC outside the SFP and HI-TRAC in the FSBs (both IP2 and IP3), including theinformation described below, and provide adequately detailed operations descriptions inChapter 10 for addressing these conditions. (CSDAB)The STC function is like that of the transfer cask that many spent fuel storage systems use toload fuel from the pool and transfer it to the storage overpack. Modifications made to thelicensing report in response to staff RAIs indicate that the STC dose rates (representativeloading) are significantly higher than have been analyzed for nearly all spent fuel transferdevices currently approved under 10 CFR Part 72. Given the STC's high dose rates, off-normalevents, such as crane hang-ups, may result in conditions that would not be encountered werethe crane hang-up to occur with a standard transfer cask. Thus, the applicant should providedose evaluations for personnel involved in performing operations to recover from a crane hang-up, including manual crane operations and crane repair. These evaluations should includedescriptions of personnel actions during these operations, personnel numbers and locationsrelative to the STC, duration of each operation segment, and appropriate justification for eachaspect of the evaluation. Additionally, the applicant should provide dose evaluations for otherplant personnel (e.g., administrative staff, guards, plant technicians, etc.) and describe theimpacts this event would have on operations/activities in site facilities adjacent to, or near therespective units' FSBs. Evaluations should also be performed for members of the public on-sitePage 50 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONand at the controlled area boundary, with adequately justified bases and assumptions. Chapter10 of the licensing report should provide an adequately detailed description of these operations,with which the evaluations in Chapter 7 should be consistent.This information is needed to confirm compliance with 10 CFR 20.1101(b) and 20.1301(a) and(b), GDC 61, and the intent of 10 CFR 72.104 and 72.126.Response to RAI 7-4A dose evaluation for the off-normal condition of a crane hang-up has been performed. Notethat this crane hang up is the only off-normal condition during the STC movement from thespent fuel pool to the HI-TRAC. The following have been addressed in Chapters 7 and 10 of thelicensing report. All operations will be performed following Indian Point's radiation protectionprogram which will ensure that ALARA objectives are met.Dose to the Operator Evaluation: The dose rates to the primary and secondary operatorsin case of crane hang up are documented in Table 7.4.6. Two locations are consideredin this respect, one directly above the STC lid (8 feet from the STC lid) for manual craneoperation and the other one 22 feet from the surface of the STC for supervising themanual crane operation. Note that the 8 feet distance represents the minimum approachdistance to the STC and the distance will increase as the crane lowers the STC into theSFP or HI-TRAC. Dose rates (mrem/hr) are reported in Table 7.4.6, so that Indian PointRadiation Protection personnel can estimate the person-rem depending on the particularsituation. It is not expected that more than 3 personnel will be needed for maneuveringthe crane manually. Two operators will be stationed at the sides of the trolley tomanually release the brakes and one will observe the load from a distance. Four hoursduration is considered and used for the annual dose calculations in case of crane hangup. Four hours is sufficient time to either return the STC to the SFP or place it into theHI-TRAC by manually operating the crane. Dose to the operator from the crane hang upis discussed in Section 7.4.4, while the operational part is described in Section 10.5.1.Table 7.4.6 demonstrates that the dose to the operators in case of crane hang up will bereasonable considering the short duration of the event.* Other Plant Personnel: Note that only plant personnel required to perform the wettransfer operation will be allowed near or inside the FSB during STC movement from theSFP to the HI-TRAC.Dose to the Public (On-site): Dose to the public is discussed in Section 7.4.5 for normalSTC movement between the SFP to the HI-TRAC and also in case of crane hang up.The dose rates to the member of public during normal operation are documented inTable 7.4.7, and Table 7.4.8 demonstrates the dose rates for crane hang up meetregulatory limits. A 60 m distance is required for the member of the public from thesurface of the STC. Conservatively, the same 60 m distance around the FSBs at Unit 3and 2 will be imposed as restricted area and will be controlled in accordance with theTechnical Specifications. Note that the fuel storage building is not credited for thiscalculation.* Impact in Adjacent Facilities: As a 60 m restricted area from the FSBs is required fornormal operation and will be controlled in accordance with the Technical Specifications,the crane hang up event will have minimal impact on the adjacent facilities. In otherPage 51 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONwords, all the adjacent buildings within 60 m radius from the FSBs will be restricted toother plant personnel and member of public during the STC movement from SFP to HI-TRAC. Note that 60 m is conservative as this is calculated without taking credit of any ofthe existing surrounding structures.NRC RAI 7-5The following editorial errors were identified. Provide corrected SAR pages for review.(CSDAB)a. Table 7.4.14 should list the regulatory accident dose limit units as mrem and notmrem/yr. Accident dose limits in 10 CFR 72.106(b) are given in terms of dose and notdose per year.b. The first paragraph in Section 7.0 should be modified to clearly indicate that off-normalconditions are limited per 10 CFR 72.104 together with normal conditions. The currentparagraph text appears to incorrectly indicate that off-normal conditions are limited withaccident conditions by 10 CFR 72.106 limits.c. Ensure appropriate terms are used for the respective dose evaluations. The current textindicates that site boundary dose rates are used to show compliance with the intent of10 CFR Part 72 dose limits. Dose limits in 10 CFR Part 72 are for the controlled areaboundary, as defined in that Part. While compliance with the intent of Part 72 limits isattempted to demonstrate compliance with 10 CFR Part 50 and Part 20 limits, the term'site boundary' is a 10 CFR Part 20 term. The relationship between the controlled areaboundary and the site boundary can be elaborated to justify how demonstration ofcompliance with dose limits for the one equates to demonstration of compliance with thedose limits for the other.This information is needed to ensure compliance with 10 CFR 20.1301(a), 10 CFR 50.36a andthe intent of 10 CFR 72.104 and 72.106.Response to RAI 7-5The editorial errors in the licensing report, as noted in the RAI, have been addressed as follows.a. The accident dose limit units have been corrected and mrem is specified instead ofmrem/year in the appropriate Chapter 7 tables (Tables 7.4.18 to 7.4.20).b. Chapter 7 has been extensively revised. In Section 7.1.3 it is identified that the off-normal condition dose limits are in accordance with 10 CFR 72.104.In revised Chapter 7 the 10 CFR Part 72 term "controlled area boundary" is used instead of "siteboundary" when demonstrating compliance to Part 72. Controlled area boundary andaccompanying regulations are discussed in Section 3.1.2.NRC RAI 7-6Provide an occupational dose assessment that captures all elements of the spent fuel transferoperations and include appropriate justification of assumptions regarding personnel numbersand locations relative to the STC, time durations, applicable dose rates, and adequacy ofassessment detail and description accuracy. Additionally, an inspection found that during liftsof the STC, there is a 13/16" gap between the lid and the flange; the assessment shouldconsider this gap and STC operations should be modified, as appropriate, to account for thisPage 52 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONgap. (CSDAB)Accurate and adequate assessments of occupational dose are important to ensure appropriateconsiderations are taken in the development of detailed procedures and the planning ofoperations. Staff continues to have concerns regarding the occupational dose assessmentprovided in the licensing report. Staff's initial concerns (see first round RAI questions 7-17 and7-18) were only partly addressed, and further questions have arisen due to changes made tothe assessment and the much higher dose rate estimates for the loaded STC. First, it is notclear that the assessment includes all elements of the operations. A comparison with theassessments done for the HI-STORM 100 system's loading operations illustrates the level ofdetail and comprehensiveness that would be expected, allowing for differences betweensystems (e.g., bolted closure vs. welded closure). Additionally, new procedures have beenintroduced, such as the 24-hour pressure rise test. Second, there continue to beinconsistencies between the assessment's description of operation conditions and thoseanalyzed in the shielding models and the descriptions in Chapter 10 of the licensing report.Third, dose rates for some operations appear to be inappropriate for the actual personnellocations and configurations as understood from the Chapter 10 descriptions. Examples includeoperations on the STC lid using dose rates from the HI-TRAC axial side. Finally, variouschanges to the numbers and locations of personnel relative to the operations have been madewithout explanation or basis. It is not clear how fewer people than previously stated arerequired for the same operations. It is also not clear how they are to perform their functionsfrom greater distances. For example, operations for raising the STC from the SFP, such assurveying the STC lid dose rates and washing the STC and crane equipment with clean waterappear to be performed from 10 meters distance from the STC. Also, some operations of asimilar nature are done under different conditions (e.g., placing STC in HI-TRAC vs. movingSTC from HI-TRAC into SFP).This information is needed to confirm compliance with 10 CFR 20.1101(b) and the intent of 10CFR 72.104 and 72.126.Response to RAI 7-6Occupational doses to the radiation workers have been reevaluated and are reported in therevised licensing report. The results of this reevaluation are presented in Table 7.4.22. Section7.4.12 discusses the occupational dose calculation. All operations will be performed followingIndian Point's radiation protection program which will ensure that ALARA objectives are met,e.g. lead blankets will be used for radiation protection as applicable. Moreover, the operatorswill not all be at the closest location for the entire duration with respect to the STC or HI-TRACduring a specific operation. The operators will perform their stipulated tasks during a specificoperation and, on completion of that task, wait in a low dose waiting area. Therefore, anadditional column is introduced in the occupational dose table (Table 7.4.22), which documentsthe duration at closest distance with respect to the STC or HI-TRAC. These durations at theclosest distance are utilized for the occupational dose calculations.The occupational dose to the workers has been reevaluated considering the following:All the dose rates from the bare STC are evaluated assuming a 13/16" gap between thelid and the flange. A steel ring is attached with the lid to minimize streaming through thisgap. Additionally, the few localized relatively high dose locations surrounding STC, suchPage 53 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONas on surface of the ribs, are clearly shown in the licensing report and the Indian Pointradiation protection personnel will develop their ALARA plans accordingly.* All elements of the operations as presented in Table 7.4.22 are covered with as muchdetail as practical. The occupational dose calculation steps are consistent with theoperation description in Chapter 10 of the licensing report.* All the required procedures performed by the worker in the radiation zone are accountedfor in this new occupational dose evaluation, including the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months pressure rise test." Appropriate dose rates are used for the dose to the worker calculations. For example thedose rate from the outer ring of the STC top lid (from Table 7.4.6) is used for all theoperations on the STC lid. Note that dose rates are calculated without the HI-TRAC toplid for HI-TRAC related operations. Workers will perform the HI-TRAC top lid operationsfrom the side of the HI-TRAC by extending their hands and other equipment.Nevertheless, for HI-TRAC lid operations dose rates from the outer ring of the STC lid isutilized.* The locations and number of personnel involved have been revised as detailed in thelicensing report. These inputs to the dose analysis are based on a careful evaluation ofthe expected operational procedures and other spent fuel transfer experiences." Remote operational tools will be employed as far as practicable to minimize the time thata worker is near the STC.It can be concluded from Table 7.4.22, that the occupational dose associated with fuel transferoperations employing the STC/HI-TRAC is reasonable.NRC RAI 7-7Explain whether normal operations with the loaded STC by itself are done remotely, includinghow remote operations are performed, the kinds of equipment used, and the quality standardsemployed for such equipment. If normal conditions are dependent upon remote operations, anappropriate condition should be included in the proposed TS, Appendix C. (CSDAB)With revised STC dose rates much higher than has been evaluated for normal spent fuelloading or unloading, for currently approved spent fuel storage systems and the revisions toworker positions relative to the STC outside the SFP and the HI-TRAC, it is not clear whether ornot operations are expected to be done in a remote fashion such that under normal conditionspersonnel are not around the STC like they are for loading operations for approved spent fuelstorage systems. If the evaluation and operations rely upon remote operations, the applicationshould clearly indicate that is the case, providing a description of how remote operations are tobe performed (e.g., optical guidance systems and remote crane maneuvering), an explanationof how some steps can be performed remotely, and the assurance of equipment reliability forsuch operations. Chapter 10 of the licensing report should also be modified to clearly indicatethat operations are performed remotely. Additionally, a TS condition should be added to statethat the STC is handled remotely when out of the SFP and the HI-TRAC in conjunction withPage 54 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONappropriate ALARA practices and that equipment assigned appropriate quality standards forremote handling operations will be used for such operations, with consideration for redundancyof such equipment.This information is needed to confirm compliance with 10 CFR 20.1101(b), 10 CFR 50.34, andthe intent of 10 CFR 72.104, and 72.126.Response to RAI 7-7The estimated average dose rates at the side of the STC (right below the upper flange) are lessthan 4 rem/hr similar to the estimated average dose rates at the side of the Holtec HI-TRACtransfer cask (from HI-STORM FSAR) when used with MPC-32 (3.6 rem/hr). It is noted that,local dose rates at the rib locations on the sides of the STC are higher than the dose rates at therib locations of the Holtec HI-TRAC transfer cask, however, these dose rates do not exceedthose for other normal reactor maintenance operations.In keeping with ALARA practices, operator dose will be minimized through a combination ofmaximizing distance from the source and limiting the operator time in the dose field. Thehighest dose rates that occur during the fuel transfer evolution take place when the STC is inopen air when traveling between the spent fuel pool and the HI-TRAC. During these operations,the crane operator will use a wireless radio controlled pendant to operate the crane so that theoperator can be at a distance from the STC while still maintaining visual contact. The operatorwill be on the side of the pool/truck bay opposite of the STC which is a minimum distance ofabout 8 meters. The crane spotter will use remote cameras to help guide the operation wheninserting the STC into the HI-TRAC and into the spent fuel pool. He will be located in a positionsimilar to that of the crane operator.The bottom of the STC has a generous taper and the top of the STC centering assembly has alead-in with a 45 degree angle and 2" width to facilitate aligning the STC with the opening in theHI-TRAC. There is no need for additional operations personnel to be near the HI-TRAC whenthe STC is being installed. During unloading, the clearance of the sides of the STC to the spentfuel pool racks, walls, and tool holders is at least 8", therefore the STC can be positioned in thepool and lowered into the water to reduce dose before any final adjustments in position arerequired. The positioning of the crane can be further enhanced using alignment aids such aslaser pointers mounted to the crane bridge and/or trolley along with targets on the fuel buildingfloor to repeatedly locate the STC in the HI-TRAC and spent fuel pool without any significantintervention or adjustment of the crane position.The wireless control for the crane is a commercial item that is in use at most nuclear plants inthe US. They are used at many sites that load the Holtec HI-STORM system. The system hasbeen proven reliable and will be tested during the required dry runs. The most common causefor problems with the remote controls is low battery power and interference from other radioequipment in the area. Spare batteries will be maintained during loading and radio interferenceissues are controlled through the plant procedures related to RFI/EMI controls. There is nospecial testing or qualification needed for the wireless controls for the crane.NRC RAI 7-8Perform an evaluation to demonstrate compliance with 10 CFR 20.1301(b) for STC movementbetween the SFP and the HI-TRAC for both loading and unloading operations, implementingPage 55 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONadditional controls to ensure compliance with this requirement, including appropriate TS, asnecessary. (CSDAB)It is not clear if the current evaluation addresses compliance with the limits of 1OCFR 20.1301 (b)for the condition of the STC by itself, including for normal conditions. Given the high dose ratesof the STC and the relatively close proximity of other buildings where members of the publicmay be located (as seen in Figure 7.4.2 of the licensing report) while loading or unloadingoperations are ongoing, it is not clear whether additional controls are needed to ensurecompliance with 10 CFR 20.1301(b). Any assumptions regarding shielding provided by buildingmaterials and structures should be appropriately justified. Consideration should also be given tothe need for additional conditions in the proposed TS, as necessary, to ensure appropriatecontrols are instituted to ensure compliance with 10 CFR 20.1301(b).This information is needed to confirm compliance with 10 CFR 20.1301(b).Response to RAI 7-8An evaluation has been performed to demonstrate compliance with 10 CFR 20.1301(b) for theSTC movement between the SFP and the HI-TRAC and the results are presented in Tables7.4.7 and 7.4.8 of the licensing report. The evaluation shows that a 60 m restricted area isrequired from the surface of the STC. Conservatively, the same 60 m distance around the FSBsat Unit 3 and 2 will be imposed as restricted area and will be controlled in accordance with theTechnical Specifications during the STC movement from SFP to HI-TRAC. Note that the fuelstorage building itself nor any other surrounding structures are credited in this calculation.The proposed TS has been revised to include the restricted area for STC movement betweenthe SFP and the HI-TRAC.NRC RAI 7-9Provide further justification regarding the assumed cobalt impurity levels. (CSDAB)While staff accepts that the dose evaluations for the public and the controlled area boundaryuse fuel with decay times appropriate for the assumed cobalt impurity levels (as stated inresponse to first round RAI 7-14), this does not hold true for the occupational dose assessmentsthat use fuel that, based upon the decay time and the burnup, would have been manufactured ina time when cobalt was not controlled to limit its amount in assembly hardware as it is for morerecently fabricated fuel (post 1989). Additionally, it is still not clear that Non-Fuel Hardware(NFH) impurity levels may not be higher as well since the combination of proposed burnup andcooling time limits for NFH indicate that these would also have been fabricated during the periodbefore cobalt reduction efforts were begun. Thus, the applicant should provide justification forthe impurity level assumed for the fuel contents used in the occupational dose evaluations andthe NFH (in all evaluations) or modify the evaluations to account for higher cobalt levels, on parwith what has been identified in literature as the cobalt levels found in assembly hardware fromthat time (pre-1989).This information is needed to confirm compliance with 10 CFR 20.1101(b), 10 CFR 50.90 and50.34a(c) and the intent of 10 CFR 72.104 and 72.106(b).Response to RAI 7-9Page 56 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONThe revised analyses utilize maximum cobalt impurity levels, which are obtained from themeasurements of the composition of the assembly hardware used by the Indian Point unit 3assembly manufacturers (Westinghouse). The maximum cobalt-59 content of the pre-1989 fuelassemblies in the unit 3 SFP is 1.2 gm/kg of steel, whereas the post-1989 assemblies contain0.5 gm/kg (maximum) of cobalt-59. These cobalt-59 impurities are used in the shieldingcalculations. More specifically, 1.2 gm/kg is applied for assemblies with 20 years cooling (outerbasket locations) and 0.5 gm/kg are used for assemblies with 10 years cooling (inner basketlocations). Additionally, the unit 3 pool has 6 assemblies with inconel spacer grids. These areaccounted for in the calculations by assuming all 4 inner region assemblies contain inconelspacer grids. 4.7 gm/kg of cobalt-59 is assumed for inconel.Since relatively long cooling times (greater than 15 years) may be required for BPRAs andTPDs (especially for the TPDs with high burnups), the value of 1.2 gm/kg of cobalt-59 is usedfor steel in all BPRAs and TPDs. 4.7 gm/kg of cobalt-59 is applied for inconel in those devices.For RCCAs only inconel is considered with 4.7 gm/kg of cobalt impurity. NSAs are analyzedwith 0.5 gm/kg of cobalt-59 in steel as 10 years of cooling time is assumed in the doseevaluations for those devices. However, for NSA inconel is again considered with 4.7 gm/kg ofcobalt impurity.All cobalt impurity levels used in the calculations, including assembly fittings and NFH, arediscussed in Section 7.2.2 of the licensing report.NRC RAI 7-10Provide the basis for the division of off-normal conditions into two categories of time duration (8hours and 30 days) and describe the kinds of events/conditions that fall into the two differentcategories, including adequate justification. (CSDAB)It is not clear that the division of off-normal conditions into two categories of differing durations isappropriate or justified. It is also not clear what kinds of events or conditions would beconsidered to fall into one or the other of these categories nor why their classification as one orthe other type of off-normal condition would be justified. Staff notes that assumptions regardingoff-normal conditions may impact whether or not compliance with the intent of 10 CFR 72.104(a)can be demonstrated, even in cases where only a single transfer is considered as off-normaland the remaining transfers are normal; thus, evaluations to demonstrate this complianceshould be modified, as necessary.This information is needed to confirm compliance with 10 CFR 50.34 and the intent of 10 CFR72.104(a).Response to RAI 7-10In the previous submittal an 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months off-normal condition was considered for HI-TRACmovement between unit 3 and 2 with 10% fuel rod breaches. In the current submittal, thesealing system of the STC will be tested to leak tight criteria per ANSI 14.5. Therefore, aneffluent contribution to the dose rates is no longer considered in the current revision of thelicensing report, which eliminates the requirement of the off-normal condition with 10% rodbreaches.In the current revision two off-normal conditions are postulated, one for bare STC movementfrom the SFP to the HI-TRAC and the other one for the HI-TRAC movement between unit 3 andPage 57 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATION2. Crane hang up (4 hrs) is considered as the off-normal condition for STC movement betweenthe SFP and the HI-RAC. The postulated off-normal condition for the HI-TRAC is the transporterbreak down (30 days event).NRC RAI 7-11Justify the change in the hours used to determine the annual dose contribution to the controlledarea boundary of the Independent Spent Fuel Storage Installation (ISFSI) and other sitefacilities or use the originally assumed hours. (CSDAB)In its resubmittal of the licensing report along with the RAI responses, the applicant changed thenumber of hours per year used in the determination of the annual dose contribution from theISFSI and other site facilities for evaluation of annual doses against the 72.104(a) limits. Thehours were reduced to 192 from 500. While 192 hours0.00222 days 0.0533 hours 3.174603e-4 weeks 7.3056e-5 months is the total hours expected for the totalnumber of spent fuel transfers anticipated in a given year (currently taken to be 24), the ISFSIand site facilities contribute to dose at the remaining times of the year when spent fuel transfersare not occurring. Based upon previous arguments, 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months seems to be a more justifiabletime estimate to use for these facilities for 72.104(a) evaluations. As part of any justification, theapplicant should also include dose evaluations for a controlled area boundary at a distance of137 meters (based upon the distance assumed in the ISFSI 72.212 evaluation), calculating theISFSI and site facilities' dose contributions assuming 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months per year and 24 spent fueltransfers. These evaluations would need to address both normal and off-normal conditions.Staff notes that the assumptions regarding exposure time may impact the ability to demonstratecompliance with the intent of 10 CFR 72.104(a), even in cases where only a single transfer isconsidered off-normal and the remaining transfers are normal; evaluations to demonstrate thiscompliance should be modified, as necessary.This information is needed to confirm compliance with 10 CFR 50.34, 10 CFR 50.34a, and theintent of 10 CFR 72.104(a).Response to RAI 7-11In the previous submittal the time used to determine the annual dose contribution to thecontrolled area boundary of the ISFSI and other site facilities was 192 hours0.00222 days 0.0533 hours 3.174603e-4 weeks 7.3056e-5 months per year. In thecurrent submittal Chapter 7 of the licensing report has been extensively revised and 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months sper year is used together with a controlled area boundary at a distance of 137 meters from theedge of the ISFSI. This change is reflected in Tables 7.4.16, 7.4.17, 7.4.18, 7.4.19, and 7.4.20.Note that in the current submittal, 16 transfer operations between unit 3 and 2 are considered asthat is the maximum targeted number of transfers per year.NRC RAI 7-12Provide an evaluation of the impacts on dose rates around the loaded STC and the loaded HI-TRAC for contents including a neutron source assembly (NSA). (CSDAB)The current evaluation relies upon the arguments used in the HI-STORM 100 FSAR for NSAs.However, the allowable loading configuration in a HI-STORM 100 is such that a basket cell withan NSA is always completely surrounded by basket cells loaded with fuel not containing NSAson all sides. This is not the case in the STC basket. Furthermore, some NSAs havesignificantly long half-lives and source strengths similar to design-basis fuel assemblies. ThisPage 58 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONwas the basis for the restriction of only a single NSA loaded in the very center of the MPC. Inthe STC, no basket cell where a NSA may be loaded is completely surrounded by basket cellswithout NSAs. Thus a more detailed evaluation of the impacts on dose rates around the loadedSTC and the loaded HI-TRAC should be performed to show these impacts, including azimuthalvariations to capture the areas around the STC and around the HI-TRAC where the outerbasket cell contents do not shield the inner basket cells where NSAs are permitted. Theevaluations should also address any potential impacts on the doses to personnel and membersof the public.This information is needed to confirm compliance with 10 CFR 20.1101(b), 20.1301 (a) and (b),10 CFR 50.34, and the intent of 10 CFR 72.104, 72.106 and 72.126.Response to RAI 7-12An evaluation has been performed of the impacts on dose rates around the loaded STC and HI-TRAC for contents including a NSA. In the evaluation NSAs are restricted to one per cask andalso restricted to the inner region. These restrictions are incorporated into proposed TS Table3.1.2-2. In this evaluation the NSA source terms are described in Section 7.2.2 of the licensingreport and the NSAs are evaluated at 360,000 MWD/MTU burnup and 10 years cooling time.This change has also been incorporated into proposed TS Table 3.1.2-2. The dose results onthe surface, and 1 m from the STC and HI-TRAC are reported in Table 7.4.21. For side doserates from both the STC and HI-TRAC, dose rates are reported at 00 (on the rib) and at 450. The450 location is selected to show the dose rates on the surface and at 1 m, where the outerbasket cell contents does not shield the inner basket cells. Table 7.4.21 demonstrates that ingeneral one NSA and 11 BPRAs are bounded by 12 BPRAs for the side, and for the bottom theNSA is bounded by the RCCAs. Note that in some cases the dose rates from 1 NSA and 11BPRAs combination are slightly higher than that of 12 BPRAs case. Further investigation hasrevealed that the difference in the radial surface dose rates (Surface 00 and 1 m away fromsurface 00 for the bare STC) are not statistically significant, i.e. results are within one standarddeviation. Additionally, the top dose rates (bare STC) between the two cases (1 NSA and 11BPRAs vs. 12 BPRAs) are also comparable. HI-TRAC dose rates with NSA at 450 (on surface)and for the top lid are marginally higher than that with the BPRAs. This marginal difference isdue to the NSA neutron source term. However, it is important to note here that the HI-TRACdose rates are calculated without considering borated water inside the STC. Therefore, BPRAsand RCCAs are used to report the maximum doses and to show compliance with the regulatorylimits.NRC RAI 7-13Modify the accident dose evaluation to include the following configurations as a result of a tipover accident: (CSDAB)a. dose resulting from exposure of the HI-TRAC base, side, and top, accounting for anyareas of the STC basket that are no longer submerged in water as a result of the cavitywater receding from the side of the STC now facing up, andb. the STC off-center in the HI-TRAC as a result of the tip over accident and the crushing ofthe inner impact limiter/STC centering assembly, with the loaded HI-TRAC in the verticalorientation.An evaluation of the tip over accident should address applicable shielding and dose rateconditions and evaluations. According to the structural evaluation, the STC centering assemblyPage 59 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONacts as an impact limiter and crushes during the tip over accident; however, it is not clear howmuch crush will occur. The shielding evaluation should consider the case where the STC is nolonger centered in the HI-TRAC, shifting the STC from center to the extent that the structuralevaluation shows the centering assembly will crush. The assumption of loss of the water jacketshould also be used in these evaluations as it is for the current accident evaluation.This information is needed to confirm compliance with 10 CFR 50.34 and the intent of 10 CFR72.106.Response to RAI 7-13The accident dose evaluation has been modified to include the RAI identified configurations asa result of a tip over accident. These additional two conditions are:1. Simultaneous loss of water from the HI-TRAC water jacket and from the annular regionbetween the HI-TRAC and STC. Note that the loss of water from the annular regionbetween the STC and the HI-TRAC was modeled instead of water receding from theside of the STC now facing up after a tip over, since the complete loss of water from theannular region would bound the condition of relocation of water in the STC from the tipover.2. STC off-center as a result of tip over accident accompanying with simultaneous loss ofwater from water jacket and HI-TRAC annulus.Note that as in all other HI-TRAC calculations, these two conditions are also modeled withoutthe HI-TRAC top lid. The results are reported in Tables 7.4.12 and 7.4.13 of the licensing report.Additionally, these two accident conditions are evaluated to show compliance with 10 CFR72.106 at the controlled area boundary. Results of those evaluations are presented in Tables7.4.19 and 7.4.20.NRC RAI 7-14Provide an evaluation that demonstrates the bounding dose rates for a loaded STC containingthe contents permitted by the proposed TS contents limits. (CSDAB)A regionalized loading pattern is used for some dose evaluations in the application. Staff askedthe applicant in the previous RAI (see question 7-15 of the first RAI) to justify the use of theselected source terms in the regionalized loading pattern and the bounding nature of the doserates and dose estimates from these sources. The applicant discussed use of a uniform loadingpattern to perform some of the evaluations, stating that it exceeds the limits of what is allowedby the proposed decay heat limits. The applicant's response does not address the issue,especially in light of the very high dose rates from the STC with the representative loading.Decay heat limits and dose rates do not correlate on a one-to-one basis given that differentcombinations of burnup, decay time and minimum enrichment can yield the same decay heatbut quite different radiation source terms. For the proposed operations, the STC operates muchlike a transfer cask and can be considered a lightweight transfer cask, since it is the device thatis used to load and unload fuel from/into the spent fuel pools. With the regionalized loading alsoresulting in very high dose rates on the STC, much higher than has been evaluated for allapproved dry storage systems' transfer casks, it is important to understand the maximum doserates that may be obtained for the allowable contents; even small relative dose rate variationsPage 60 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONcan mean large changes in dose rates. Understanding the maximum dose rates that can occurduring transfer operations will enable proper ALARA and operations planning as well asdemonstration that transfer operations with all proposed contents will meet the regulations at allstages of the operations. Dose evaluations should be modified as necessary to account for thebounding dose rates.This information is needed to confirm compliance with 10 CFR 20.1101 (b) and 20.1301(a) and(b), 10 CFR 50.90 and 50.34a(c) and the intent of 10 CFR 72.104.Response to RAI 7-14A revised dose evaluation has been performed that has determined the bounding dose rates fora loaded STC containing the contents permitted by newly proposed TS contents limits. Therevised analysis recognizes that decay heat limits and dose rates do not correlate on a one-to-one basis given that different combinations of burnup, decay time and minimum enrichment canyield the same decay heat but quite different radiation source terms.In the revised analysis five loading combinations which represent the majority of the currentassemblies in the Indian Point unit 3 spent fuel pool were selected. The remainder of theassemblies in the spent fuel pool, and those to be added to the pool inventory during refueling,will also be covered by these selections after additional cooling in the SFP. These selections arereported in Table 7.1.1 of the licensing report and are also included in the TS. The dose ratesare evaluated for all five loading patterns and are presented in Tables 7.4.1 and 7.4.9 for STCand HI-TRAC, respectively. Table 7.4.1 show that loading pattern 4 is bounding from shieldingperspective. Hence, loading pattern 4 is utilized for all other bare STC dose calculations inChapter 7. On the other hand, Table 7.4.9 establishes that in general the loading pattern 3 isbounding for the side, while loading patterns 4 is bounding for top and bottom for the HI-TRAC.However, it is observed that this trend is not universal and depends on the dose locations andtransfer situations (normal or accident) [L.G]. In Chapter 7, dose rates for all the HI-TARC casesare reported only for the bounding loading patterns as determined.NRC RAI 7-15Refer to RAI 7-1 from the previous RAI letter. Evaluate the effect on dose and potential effecton canister leakage rate assuming a 10% fuel rod breach for all off-normal conditions. (TCB)Section 7.4.5 "Effluent Dose Evaluation" of Report HI-2094289 identifies the first off-normalcondition as a breakdown of the cask transporter without HI-TRAC recoverability for 30 days,but only assumes 1% fuel rod breach. The percent of spent fuel postulated to fail for off-normalconditions is 10% as identified in Table 5-2 of NUREG-1536 Rev. 1 "Standard Review Plan forSpent fuel Dry Storage Systems at a General License Facility". There appears to be nojustification for only assuming 1% fuel rod breach. The 10% release fraction identified in thestandard review plan is a bounding value for off-normal conditions and is not meant to bereduced based on postulated specific scenarios. Similarly, we would not expect the 100% fuelrod breach for accident conditions to be reduced based on a postulated specific accident, bututilized as a bounding value.10 CFR Part 50, Appendix B, Criterion III, Design Control states in part, that measures shall beestablished to assure that applicable regulatory requirements and the design basis .... arecorrectly translated into specifications, drawings, procedures, and instructions. Also, ASMECode,Section III, NCA-4000, Article 4134.3(a) references ASME NQA-1, where in SupplementPage 61 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATION3S-1, Section 3.1 it states that design analyses such as physics, stress, hydraulic and accident,shall be performed in a planned, controlled and correct manner.Response to RAI 7-15The effect on dose and the potential effect on canister leakage rate assuming a 10% fuel rodbreach for off-normal conditions no longer requires evaluation. A new approach has been takenand the sealing system of the STC will now be tested to the leak tight criteria per ANSI 14.5 asdescribed in the response to RAI 8-1. The factory leak testing of the entire STC confinementboundary is also discussed in that response. Therefore, the effluent contribution to the doserates is not considered in the latest revision of the licensing report.NRC RAI 7-16Refer to RAI 7-2 from the previous RAI letter. Justify or remove the statement at the end of thethird paragraph in Section 7.4.5 "Effluent Dose Evaluation" of Report HI-2094289 that fines,volatiles and crud would remain entrapped within the water environment. If applicable, submit arevised section 7.4.5 as part of your RAI response. (TCB)Since the STC, the confinement boundary, is tested to a finite leak rate (i.e. not leaktight) andcan be pressurized (refer to Figure 5.3.2 "24 Hour Pressure Rise Under Design BasisConditions," a pathway exists for the potential release of radioactive material. No credit can betaken for the HI-TRAC since it is not part of the confinement boundary. Hence, making astatement that this contamination would remain within the water seems invalid.10 CFR Part 50, Appendix B, Criterion III, Design Control states in part, that measures shall beestablished to assure that applicable regulatory requirements and the design basis .... arecorrectly translated into specifications, drawings, procedures, and instructions. Also, ASMECode,Section III, NCA-4000, Article 4134.3(a) references ASME NQA-1, where in Supplement3S-1, Section 3.1 it states that design analyses such as physics, stress, hydraulic and accident,shall be performed in a planned, controlled and correct manner.Response to RAI 7-16As discussed in the response to RAI 7-15 a new approach has been taken and the sealingsystem of the STC will now be tested to the leak tight criteria per ANSI 14.5 as described in theresponse to RAI 8-1. The factory leak testing of the entire STC confinement boundary is alsodiscussed in that response. Therefore, the effluent contribution to the dose rates is notconsidered in the latest revision of the licensing report, which additionally eliminates therequirement for HI-TRAC credit as a part of confinement boundary. The above mentionedstatement from Section 7.4.5 has been deleted from the revised licensing report.NRC RAI 7-17Refer to RAI 7-3 from the previous RAI letter. The response to NRC RAI 7-3 provided by letterdated October 5, 2010, included an unnumbered table on page 57 of Attachment 1 whichsummarizes inputs used to calculate the atmospheric dispersion factors (x/Q values) used in theeffluent dose evaluation. (AADB)a. Please explain how the uy and oz values listed in the table on page 57 were derived fromFigures 1 and 2 of Regulatory Guide (RG) 1.145, "Atmospheric Dispersion Models for PotentialPage 62 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONAccident Consequence Assessments at Nuclear Power Plants." For example, were oy and azcomputer generated or were they derived by approximation directly from the figures? Pleaseprovide additional detail, including computer generated summaries and/or annotated figures,which show how the calculations were made, particularly at 1 and 20 meters as these distancesare less than the 100 meter minimum distance plotted in Figures 1 and 2.b. Please explain the basis for why x/Q values were calculated for a distance of 1 meter.Please provide additional detail supporting the use of the RG 1.145 methodology to determinex/Q values considering that the intent of the guidance does not appear to be appropriate for thisdistance. RG 1.145, which provides guidance for calculation of x/Q values applicable to typicalexclusion area boundary and low population zone distances, implicitly assumes a minimumdistance of 100 meters with regard to information provided in Figures 1 and 2. Further, NRCstaff notes that other NRC documents such as RG 1.194, "Atmospheric Relative Concentrationsfor Control Room Radiological Habitability Assessments at Nuclear Power Plants," state thatNRC guidance for the calculation of x/Q values may not apply under certain conditions, such asat distances less than about 10 meters.c. The application references Interim Staff Guidance (ISG)-5, "Confinement Evaluation," as thebasis for the stability and wind speed inputs. ISG-5 states that the use of stability category Dand a wind speed of 5 meters per second (m/s) are acceptable for the normal and off-normalcase calculations, and stability category F and a wind speed of 1 m/s are acceptable for theaccident case calculation. These are default values for dry cask storage systems. NRC staffnotes that these values are based on a generic approximation of meteorological conditions foran average site in the United States. Please clarify whether the normal and off-normal casesare assumed to occur during 50 percentile and the accident case during 95 percentilemeteorological conditions, respectively, as is typically assumed for other nuclear reactor effluentdose evaluations. Please confirm how the use of these default values is justified whencompared with representative 50 and 95 percentile meteorological conditions at the Indian Pointsite.d. Figure 7.4.2 of Holtec International Report HI-2094289 (ADAMS Accession NumberML103080113) is a site map showing the haul path and an exclusion area boundary (EAB),which the licensee has defined for use in the current license amendment request. This EABdoes not appear to be the current Indian Point licensing basis EAB for either Unit 2 or Unit 3defined by 10 CFR Part 50.2. Therefore, please explain the relationship of the EAB shown inFigure 7.4.2 and the current Indian Point licensing basis EABs. In addition, please provide theminimum distance between any point along the haul path and 1) the Indian Point currentlicensing basis EABs, and 2) the control room intakes.This information is needed for the NRC staff's review to ensure compliance with the criteriacontained in GDC 61.Response to RAI 7-17As discussed in the response to RAI 7-15 a new approach has been taken and the sealingsystem of the STC will now be tested to the leak tight criteria per ANSI 14.5 as described in theresponse to RAI 8-1. The factory leak testing of the entire STC confinement boundary is alsodiscussed in that response. Therefore, ay, a,, x/Q, and wind conditions no longer requireexplanation or clarification as requested by RAI Parts, a, b, and c.Part d: The "exclusion area boundary" that was defined in Figure 7.4.2 of the licensing report isPage 63 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONnot related to Indian Point's exclusion area boundary. The exclusion area boundary defined inFigure 7.4.2 is temporary and only applies to STC/HI-TRAC movement between unit 3 and 2.To avoid any further confusion the term "exclusion area boundary" is removed from the reportand the term "restricted area" is introduced. Note that this is a temporary restricted area whichwill remain in effect only during the STC/HI-TRAC transfer. Two restricted area are defined, thefirst one is for the STC movement between the pool and the HI-TRAC and vice versa and thesecond one is the HI-TRAC movement between unit 3 and 2. Note that they are independent ofeach other as the two operations are not occurring at the same time.Exclusion zones for Indian Point are defined in Chapter 2 of the IP3 FSAR. The exclusion zoneboundary for Indian Point Unit 3 is at a distance of 350 m from the containment building,whereas the exclusion zone boundary for Unit 2 is defined by 520 m distance from thecontainment building. As the effluent calculations are no longer relevant to the shielding chapterof the licensing report, the locations of the control room intakes are no longer required.CHAPTER 8 -MATERIALS EVALUATION, ACCEPTANCE TESTS and MAINTENANCEPROGRAMNRC RAI 8-1Refer to RAI 8-1 from the previous RAI letter. Justify the lack of leak testing of the entireconfinement boundary of the STC as well as describe what leak tests are done in the shop andthe field. (TCB)The original RAI 8-1 included the following which was not addressed in the response or in thereferenced sections of Holtec Report HI-2094289: "Additionally, the entire confinementboundary should be leak tested in accordance with the guidance of ANSI 14.5-1997 to verifycompliance with the design leak tightness as determined in RAI 7-1, above. This leak testingshould be performed initially at the fabrication facility and periodically (within 12 months prior toeach use) to ensure that the containment leak tightness has not deteriorated over time. Theleak testing done at the time of loading fuel is usually less stringent and is done to ensure thatthe gaskets are properly seated and the containment has been assembled properly, andtypically is checked to be at least 1 E-3 ref-cm3/sec."The applicant's response stated that "Section 8.4.4 had been revised to more clearly define thefactory leakage test and the periodic leakage test of the lid gaskets that is performed duringloading operations." However upon reviewing Section 8.4.4 no distinction is made betweenfactory and periodic leak tests and no discussion is provided with regard to leak testing theentire confinement boundary.This information is needed for the NRC staff's review to ensure compliance with the criteriacontained in GDC 61 and 10 CFR Part 50, Appendix B, Criterion Xl- Test Control.Response to RAI 8-1Section 8.4.4 has been revised to discuss the factory leak testing of the STC confinementboundary. The following text has been added:The entire STC confinement boundary shall be leak tested at the factory prior to initialuse to demonstrate that the leakage rate meets the criteria for "leak-tight" as defined inANSI N14.5. The confinement boundary material and welds will be tested using the gasPage 64 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONfilled envelope (gas detector) method or other applicable method per ANSI N14.5. Theseals on the STC lid and lid cover plates shall be leak tested using the evacuatedenvelop (gas detector) method, or other suitable method as described in ANSI N14.5.The leakage rate acceptance criterion is "leak-tight" as defined in ANSI N14.5.Section 8.5.2 and Table 8.5.1 have been revised to specify that the seals on the STCconfinement boundary will undergo a periodic leakage test (within 12 months of fuel transfer) todemonstrate that the confinement boundary seals will maintain the required leakage rate of"leak-tight" and that the confinement capabilities have not deteriorated over an extended periodof use. They will also be revised to specify that the seals on the STC confinement boundary willbe subject to a pre-transfer leakage test (analogous to the "pre-shipment" test in ANSI N14.5) toconfirm that the confinement system has been properly assembled for fuel transfer. The STCseals will be leak tested to demonstrate that there is no leakage when tested to a sensitivity of 1x 10-3 ref. cc/sec in accordance with the ANSI N14.5 requirements. In addition, text has beenadded to Section 8.5.2 to define the leakage testing requirements following any maintenanceactivity that may affect the function of the confinement boundary. Retesting of the body of theSTC is only required if maintenance activities have been performed that may affect theconfinement boundary (e.g., weld repairs to confinement boundary welds). The followingrevised text has been added to section 8.5.2:The seals on the STC lid and lid cover plates shall be tested at a frequency defined inTable 8.5.1. The seals shall undergo a periodic leakage test to confirm that theconfinement capabilities have not deteriorated over an extended period of use. Theperiodic leakage test of the seals shall be performed using the evacuated envelop (gasdetector) method, or other suitable method as described in ANSI N14.5. The leakagerate acceptance criterion is "leak-tight" as defined in ANSI N14.5. The seals shallundergo a pre-transfer leakage test to confirm that the confinement system has beenproperly assembled for fuel transfer. The pre-transfer leakage test of the seals shall beperformed using the gas pressure rise method, or other suitable method as described inANSI N14.5. The acceptance criteria is no detected leakage when tested to a sensitivityof 1 x 10-3 ref. cc/sec. Failure to achieve a leakage rate below the required value shallbe cause for seal replacement, seating surface repair, or other repair and retest of theseal joint as applicable.The STC confinement boundary components shall be leakage tested following anymaintenance repair which may affect the confinement boundary function to demonstratethat the leakage rate is "leak-tight" as defined in ANSI N14.5. The confinementboundary material and welds will be tested using the gas filled envelope (gas detector)method or other applicable method per ANSI N14.5. The seals and sealing surfaces onthe STC lid and lid cover plates shall be leak tested using the evacuated envelop-gasdetector method, or other suitable method as described in ANSI N14.5. The leakagerate acceptance criterion is "leak-tight" as defined in ANSI N14.5.Table 8.5.1 entries have been revised as follows:Task FrequencyPeriodic Leakage Test of STC (lid and lid Within 12 months prior to each fuel transfer.cover plates) seals Acceptance criteria is "leak-tight" as defined inANSI N14.5.Pre-transfer Leakage Test of STC (lid and Following each fuel loading, prior to fuelPage 65 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONlid cover plates) seals transfer. Acceptance criteria for each seal isno detected leakage when tested to asensitivity of 1.0 x 10-3 ref. cc/sec.Maintenance Leakage Test of STC (lid and Following seal replacement. Acceptancelid cover plates) seals criterion is "leak-tight" as defined in ANSIN14.5.Leakage Test of HI-TRAC top lid seals. Following each fuel loading, prior to fueltransfer. Acceptance criteria is no detectedleakage when tested to a sensitivity of 1.0 x10.3 ref. cc/sec.NRC RAI 8-2Refer to RAI 8-3 from the previous RAI letter. Justify the response provided in amendedSection 8.4.3. If applicable submit a revised Section 8.4.3 as part of your RAI response. Thestaff finds that the Code required pressure test (125% of design pressure) alone is not sufficientto pressure test the bottom flange area of the HI-TRAC since it neglects the weight of the fullyloaded STC which sets on the bottom HI-TRAC flange. The applicant's submittal states thatonly the 125% of design pressure is required for the HI-TRAC pressure test, which is notcorrect. Also, indicate in this section the total test pressure needed to meet the Code requiredtest including the weight of a fully loaded STC which weighs about 40 tons. As previouslystated, a dead load equivalent to a fully loaded STC may be placed inside the HI-TRAC for thepressure test, with the weight of the HI-TRAC supported only by its trunnions, in lieu ofincreasing the pressure above 125% of the design pressure. (TCB)The applicant's response to the previous RAI stated that changes were made to address thisissue to Sections 10.1.2 "STC Preparation and Setup- HI-TRAC Inspections and Checklist" and8.5.2 "Leakage Tests". However, the staff could find no such changes made relating to thisissue. From reading the revised SAR sections, the staff is concerned that the applicant may notfully appreciate the purpose of the pressure test. The pressure test is a structural integrity testperformed before use to assure proper fabrication of the HI-TRAC. It is not a leakage test, perse, but uses the absence of leaks as its acceptance criteria (as mandated by the Code) toensure the as-constructed vessel's suitability for the loadings. Therefore more than just thegasketed area of the pool lid needs to be checked for leaks.The NRC staff also needs confirmation of the structural integrity testing planned for the STC. Ifit will only be by pressure test, demonstrate that the pressure test exceeds all deadweight loadswith appropriate margin and that the STC will be supported only by its crane attachment pointsduring the test.10 CFR Part 50, Appendix B, Criterion Xl- Test Control requires, in part, that a test programshall be established to assure that all testing required to demonstrate that structures, systems,and components will perform satisfactorily in service is identified and performed in accordancewith written test procedures which incorporate the requirements and acceptance limitscontained in applicable design documents.Response to RAI 8-2Text clarifying the requirements to account for the weight of the loaded STC on the HI-TRACPool Lid joint was added in the previous licensing report (Revision 3) in subsection 10.1.2, stepPage 66 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATION9 referring the user back to section 8.5.2 which describes the required maintenance leakagetesting for use of the HI-TRAC. However, the text describing the requirements for testing wasadded to the maintenance leakage testing in subsection 8.5.2 and not to the initial acceptancetesting described in subsection 8.4.3. The following text has been added to 8.4.3 to make thedescription complete in response to the RAI.When performing the test, the HI-TRAC shall be supported by the trunnions. TheHI-TRAC shall be loaded with a full weight STC or STC mock-up and filled with water tofully load the pool lid bolting and bottom flange welds. As an alternative to using the fullweight STC, the test pressure may be increased to account for the missing weight. Theacceptance criteria for the pressure test shall include no visible water leakage from thepool seal and drain plug as well as the rest of the HI-TRAC water boundary. All joints,connections, and regions of high stress such as regions around openings and thicknesstransition sections shall be examined for leakage.The STC pressure boundary test does not require that any dead loads are included in thetesting. The STC is lifted by the crane attached to the lid with the fuel and STC weight beingtransferred from the STC trunnions to the lid through the STC Lifting Device. The lid bolts arenot engaged during lifting such that the STC will only be lifted with the internal cavity vented toatmosphere. Therefore, a combination of internal pressure and dead weight of the fuel will notbe present in combination. The internal cavity of the STC will only be pressurized with the STCresting on the bottom of the HI-TRAC and dead weights due to lifting are not a concern. Thestructural effects on the pressure boundary that result from lifting the STC by the trunnions willbe determined by the 300% load test of the trunnions. However, as a defense in depth, theSTC hydrostatic pressure test will be conducted with the STC supported by either the top lid orthe lifting trunnions to maximize the loads on the confinement boundary welds. Text has beenadded to subsection 8.4.3 to identify this commitment.NRC RAI 8-3Make the following modifications to provide additional clarity and consistency. (CSDAB)a. Modify Section 8.4.1 to also state that the STC will be assembled in accordance withand verified to meet the TS requirements for the STC design. In addition to the licensingdrawings, the TS contain design requirements with which the fabricated STC mustcomply.b. Modify Section 8.4.5 to also state that the lead sheet will be layered so that the minimumtotal thickness meets the TS requirements for the STC design. In addition to thelicensing drawings, the TS contain a design requirement on the minimum thickness ofthe STC lead shielding.c. Modify Section 8.4.5 to include more of the response to RAI question 8-11 from theprevious RAI letter, particularly that the layering of lead sheets where each layer is madeof multiple sections will be done so that section edges in adjacent layers are offset toeliminate potential streaming paths. This aspect of fabrication is important with regard toradiation protection, and based on the response to RAI question 8-11, should have beenincluded in the referenced section of the licensing report.d. Describe the acceptance testing that will be used to ensure that areas packed with leadwool will perform in a manner comparable with the lead sheet for shielding purposes.This information is needed to confirm compliance with 10 CFR 50.34 and the intent of 10 CFR72.44(c)(4) and 72.126(a).Page 67 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONResponse to RAI 8-3a. Section 8.4.1 has been modified to address the comment.The following text has been added to the first sentence of Section 8.4.1:"The STC shall be assembled in accordance with the licensing drawings supplied in Section 1.5and the applicable STC Technical Specifications."b. Section 8.4.5 has been modified to address the comment.The following text has been added to Section 8.4.5:"The lead sheet will be layered for a minimum total thickness as specified on the licensingdrawings found in Section 1.5 and the applicable STC Technical Specifications."c. Section 8.4.5 has been modified to address the comment.The following text has been added to Section 8.4.5:"If multiple sections are used to make a layer, they are butted up tight against one another andany gaps are filled with lead wool as described above. If multiple sections are used to make thelayers, the joints between the sections are staggered to eliminate any potential streamingpaths."d. Section 8.4.5 has been modified to address the comment.The lead sheets were installed in the STC such that there will not be any gaps between leadsheets or the lead and the steel cavity which could lead to significant streaming. Any gaps thatwere observed during the lead installation were filled with lead wool which is compressed inplace to fill the void. Nonetheless, as an additional defense in depth, the STC will be subjectedto a gamma scan to demonstrate that the lead cavity is free of voids that would lead tosignificant streamingThe following text has been added to Section 8.4.5 to address the testing of the lead:"The effectiveness of the lead installation in the STC body shall be verified after fabrication byperforming a gamma scan on the accessible surfaces of the canister in the lead shieldingregion. The purpose of the gamma scan test is to demonstrate that the lead shielding is freefrom voids that may result in streaming paths through the lead. Measurements shall be takenon a 6-inch by 6-inch (nominal) grid pattern over the surfaces to be scanned. Any gamma doserates that vary significantly from the average gamma dose measurements, accounting for thepresence of the STC ribs as applicable, shall be evaluated by Holtec to determine the effect onthe dose calculations provided for the STC. Should the measured calculations using themeasured gamma dose rates show that the calculated dose rates will exceed the dose ratesused to license the STC, corrective actions should be taken, if practicable, and the testing re-performed until successful results are achieved. If physical corrective actions are notpracticable, the degraded condition may be dispositioned with a written evaluation inaccordance with applicable procedures to determine the acceptability of the STC for service.Gamma scanning shall be performed in accordance with written and approved procedures.Measurements shall be documented and shall become part of the quality documentationPage 68 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONpackage."NRC RAI 8-4Refer to RAI 8-8 from the previous RAI letter. Identify and reconcile the discrepancies inapplicable rules for construction between ASME Code,Section III, Division 1, Subsection NDand Subsection NC for the construction of the Shielded Transfer Canister (STC) to ensure thatthe STC is constructed to acceptable quality standards. (SMMB)Spent fuel canisters (i.e. the confinement boundary) are normally constructed to ASME CodeSubsection NB or NC (reference NUREG-1536 section 3.4.1). The applicant is proposing toconstruct the STC to Subsection ND, as indicated in Section 1.3.1 of HI-2094289, Rev. 3. Thisapproach does not appear to provide the same degree of quality for a spent fuel storage ortransfer canister.This information is required for compliance with 10 CFR 50, Appendix A, GDC 61, and 10 CFRPart 72.122(a).Response to RAI 8-4The STC has been designed and manufactured to quality standards that are commensuratewith those of an MPC even though the STC is used only for "short term operations" (in theparlance of 1 0CFR Part 72) and is not subject to the thermal and pressure transients thattypically act on an MPC during its on-site storage and off-site transport. It is true that thereference code for the STC is ASME Section III, Division 1, Subsection ND, but, as summarizedbelow, the design and manufacturing of the equipment far exceeded the requirements of"ND", as well as "NC" of the Code, in many significant respects. The reasons for selectingSubsection ND as the reference code for the STC, as opposed to Subsections NB or NC, arediscussed in the response to RAI 8-8 from the October 5, 2010 Entergy letter (ADAMSAccession No. ML1 02910511). The key reasons are:" the STC serves a similar role to that of the HI-TRAC in that they are both fuel transferdevices (not long term storage devices); since the code of reference for the HI-TRAC isASME Section III, Subsection NF for Class 3 structures, the pressure vessel counterpartto "NF Class 3" -Subsection ND -is used as the reference code for the STC." cyclic fatigue, a core concern of Subsection NB and Subsection NC (NC-3200), is not acredible source of failure for the STC because significant thermal transients areessentially absent from the STC and the number of mechanical loading cycles isrelatively small.In what follows, the discrepancies in applicable rules for construction between ASMESubsection ND and Subsection NC for the construction of the STC are identified and evaluated.First it is noted that Articles NC-3000 and ND-3000 are essentially the same, except for thefollowing:a) Subsection NC provides an alternate set of "design-by-analysis" rules (NC-3200), whichuses allowable stress intensities (versus the "design-by-formula" method of NC-3300/ND-3300 which use allowable stresses);Page 69 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONb) NC-3300 expects the pressure vessel Category A, B and C welds to be 100 percentradiographed.The allowable stress intensities of NC-3200 are greater than the allowable stresses of NC-3300/ND-3300 and, therefore, enable the designer to reduce the thickness of the vessel. TheSTC, however, is conservatively designed to meet the lower allowable stress limits of NC-3300/ND-3300 with significant additional margins of safety. A photo of the fabricated STC isshown in Figure 8-4.1 at the end of this RAI response.Listed below are the principal design and manufacturing features of the STC that substantiatethe degree of quality for the STC confinement boundary.1) The primary membrane stress in the STC shell due to accident internal pressure plusnormal handling is only 2,054 psi as compared to the Level A allowable membranestress limit of 20,000 psi per NC-3300/ND-3300. Hence, the STC shell has a factor ofsafety of nearly 10 against primary stress failure. Under the rules of NC-3200, the LevelA allowable membrane stress intensity is 22,400 psi versus a calculated value of 2,144psi for the same load conditions. Thus, the factor of safety associated with the STC shellis even larger according to the alternate rules of NC-3200.2) The junction of the base plate and the STC shell has been made into a butt welded jointwhich exceeds the requirement of "NC" to allow the joint to be more easily examined.Indeed, the MPCs used in the industry use a corner joint (rather than the more robustand more easily examined butt weld joint). In this respect the STC exceeds the state-of-the-art MPC designs used in the industry.3) The STC shell to the top flange joint, designated in the Code (NC and other codesubsections) as a Category C joint and permitted to be made as a corner joint, has beenmade as a butt welded joint. In this respect also, the STC exceeds the MPC with respectto structural and confinement robustness.4) To provide an additional degree of quality relative to the radiography requirements ofND, the STC shell thickness is sized to provide safety margins that are nearly an orderof magnitude above code allowable for Subsection NC and ND and will be subjected toextensive radiography beyond the requirements of Subsection ND, but less than the100% radiography required by Subsection NC. The STC confinement boundary weldswere subject to radiographic examination exceeding the minimum NDE requirements ofND-5200 (i.e., spot radiography). Specifically, more than 30 radiographs were taken forthe STC confinement boundary to establish the soundness of its weld seams Allconfinement boundary welds (confinement shell longitudinal and circumferential welds,shell to top flange weld, and shell to base plate weld) have been subjected to RTexamination over all or a portion of their length.5) To provide added confidence, the STC has been helium leak tested to the leak tightcriteria of ANSI N 14.5 in the manner of a transport cask for off-site shipment. This testwas performed after the hydrostatic pressure test of the STC pressure boundary(minimum test pressure of 125% of design pressure) to provide additional confirmationthat that no thru-wall structural defects exist in the confinement boundary.Page 70 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATION6) The STC has been designed and manufactured as a Safety Related component underHoltec's Quality Assurance program (equivalent to Important-to-Safety Category A underPart 72) which is on par with the MPC and is the highest safety class that canbe assigned to a confinement vessel. No exceptions have been taken to the SafetyRelated component requirements under 1OCFR50 Appendix B.Of course, the STC, like the MPC, cannot be stamped as a Code vessel because of itsfundamental anatomical differences from a classical pressure vessel. However, as the abovedesign features make it clear, the STC has been engineered and manufactured to qualitystandards that emulate "NC" of Code, and exceed its provisions in several material respects.Hence, the degree of quality for the STC is no less than that of a spent fuel storage or transfercanister constructed to ASME Subsection NC.Finally, as asked in the RAI, the table provides the exceptions to "NC" that are applicable to theSTC confinement boundary. This table has been prepared in the same format and manner asthat used for the MPC in the part 72 dockets 72-1008 and 72-1014. Based upon the informationprovided above as well as the following table, Entergy concludes that the STC has beendesigned, fabricated, and inspected to a degree of quality that exceeds both Subsection NC andND, and in most respects, that of the Holtec MPC used for both storage and transport of spentfuel,Page 71 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONLIST OF ASME CODE EXCEPTIONS FOR STC CONFINEMENT BOUNDARYException, Justifidation &Compensatory MeasuresComparableto MPCunder docket72-1014MeetsNCMeetsND4-STCConfinementBoundarySubsection NCAGeneralRequirements.Requirespreparation of aDesignSpecification,Design Report,OverpressureProtection Report,Certification ofConstructionReport, DataReport, and otheradministrativecontrols for anASME Codestamped vessel.Because the STC is not an ASME Codestamped vessel, none of the specifications,reports, certificates, or other generalrequirements specified by NCA are required. Inlieu of a Design Specification and DesignReport, the STC licensing report, HI-2094289includes the design criteria, service conditions,and load combinations for the design andoperation of the STC as well as the results ofthe stress analyses to demonstrate thatapplicable Code stress limits are met.Additionally, the fabricator is not required tohave an ASME-certified QA program. All safetyrelated activities are governed by theNRC-approved Holtec QA program.Because the STC is not certified to the Code,the terms "Certificate Holder" and "Inspector"are not germane to the manufacturing of NRC-certified components. To eliminate ambiguity,the responsibilities assigned to the CertificateHolder in the various articles of SubsectionsND, of the Code, as applicable, shall beinterpreted to apply to the Entergy (and byextension, to the component fabricator) if therequirement must be fulfilled. The Code term"Inspector" means the QA/QC personnel of theEntergy and its vendors assigned to overseeand inspect the manufacturing process.YesNoNo£ 5- 1 4. .5. 5Page 72 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONReference Comparable Meets MeetsCmoet AMCoeCode Exception, Justification & to MPC NC NDComponent ASME Code t P C NSection/Article Requirement Compensatory Measures under docket72-1014STC NC/ND-1i000 Statement of Cask confinement boundary is designed, and Yes No NoConfinement requirements for will be fabricated in accordance with ASMEBoundary Code stamping of Code,Section III, Subsection ND to thecomponents. maximum practical extent, but Code stamping isnot required.STC NC/ND-2000 Requires Holtec approved suppliers will supply materials Yes No NoConfinement materials to be with CMTRs per ND-2000.Boundary supplied byASME-approvedmaterial supplier.STC NC/ND-2300 Provides impact The STC confinement boundary materials have No No YesConfinement testing been impact tested with satisfactory results perBoundary requirements for ND-2300. The test results also exceed thematerials, required C, values per NC-2300, except for thelateral expansion of the STC bolting material(SA-564 630 H 1100). It is noted, however, thatthe impact testing of the STC bolting materialhas been conducted at 0°F, which issubstantially below the Lowest Service MetalTemperature of the STC closure lid studs. Theminimum metal temperature of the STC closurelid studs is bounded by the freezingtemperature of the water inside the HI-TRAC. Amore realistic temperature estimate for the STCclosure lid studs during fuel transfer is at least1 00°F. The increased service temperature ofthe STC closure lid studs, coupled with the lowmagnitude of stress in the studs (SF > 9),virtually eliminates the risk of brittle fracture.Page 73 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONReference Comparable Meets MeetsComponent ASME Code Code Exception, Justification & to MPC NC NDSectionlArticle Requirement Compensatory Measures under docket72-1014STC NC-3300/NC- Requires full STC confinement boundary welds have been No No YesConfinement 5200 radiography of examined in accordance with ND-5200, whichBoundary Category A and B has provisions for partial and spot radiography.welded joints, and The use of ND-5200 is justified based on theCategory C full large factor of safety associated with the STCpenetration butt confinement shell (SF > 9). As a compensatorywelded joints, measure, the STC confinement boundary hasbeen helium leak tested to the leak tight criteriaof ANSI N14.5.STC NC/ND-4220 Requires certain The cylindricity measurements on the rolled Yes No NoConfinement forming tolerances shells are not specifically recorded in the shopBoundary to be met for travelers, as would be the case for a Code-cylindrical, stamped pressure vessel. Rather, theconical, or requirements on inter-component clearancesspherical shells (such as the STC-to-HI-TRAC) are guaranteedof a vessel, through fixture-controlledmanufacturing. The fabrication specificationand shop procedures ensure that alldimensional design objectives, includingintercomponent annular clearances aresatisfied. The dimensions required to be met infabrication are chosen to meet the functionalrequirements of the dry storage components.Thus, although the post-forming Codecylindricity requirementsare not evaluated for compliance directly, theyare indirectly satisfied (actually exceeded) in thefinal manufactured componentsSTC NC/ND-7000 Vessels are No overpressure protection is provided. Yes No NoConfinement required to have Function of cask vessel is as a radionuclideBoundary overpressure confinement boundary under normal andprotection. hypothetical accident conditions. Cask isdesigned to withstand maximum internalpressure and maximum accident temperatures.Page 74 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONReference Comparable Meets MeetsReferenCe Code Exception, Justification & to MPC NC NDComponent ASME Code RSectionlArticle Requirement Compensatory Measures under docket72-1014STC NC/ND-8000 States STC to be marked and identified in accordance Yes No NoConfinement requirement for with the drawing. Code stamping is notBoundary name, stamping required. QA data package prepared inand reports per accordance with Holtec's approved QANCA-8000 program.Page 75 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONFigure 8-4.1 Completed Shielded Transfer Canister at Holtec ManufacturingDivisionCHAPTER 10 -OPERATING PROCEDURESNRC RAI 10-1Regarding the operations descriptions, provide the following: (CSDAB)a. Justification for removing water from the STC/HI-TRAC annulus at the currentlydescribed step in the unloading procedures. Considering the step in the loadingprocedures in which the annulus water level is raised (just after the STC lidbolting is tightened to the required levels) and the operations descriptionsfollowing the water drain down for the unloading operations, it seems moreappropriate and in keeping with ALARA to not drain down the annulus water untiljust prior to loosening the STC lid bolts.b. Modification of the operations descriptions to include the steps for installation andremoval of the Bottom Missile Shield (BMS). The responses to the first roundRAI questions 1-2 and 7-6, indicate that the BMS is to be installed with the HI-TRAC empty; however, Chapter 10 descriptions need to be updated to reflectPage 76 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONthese operations and not just a check that the BMS is installed at the time theloaded HI-TRAC is to be moved.c. Modification of the operations descriptions to account for the 24-hour pressurerise check. Other evaluations, such as occupational dose estimates, should beupdated as necessary.d. Modification of Section 10.5.5 of the report to include descriptions of steps to betaken in the event gas sampling indicates damage to assemblies. These stepswould include such things as whether or not fuel is off-loaded and how it ishandled.e. Inclusion of the step for filling the HI-TRAC neutron shield.f. Descriptions of how the presence of a walkway crossing over the haul path isaddressed. Clarify whether and how the EAB also covers this walkway.g. Explanation of the kinds of delays envisioned for Section 10.5.2 and theconfigurations that are considered (e.g., whether the STC is always in the HI-TRAC or the STC may be outside of the HI-TRAC). Evaluations of theseconfigurations, as necessary, should also be provided.This information is needed to confirm compliance with 10 CFR 20.1101(b) and20.1301(a) and (b), 10 CFR 50.34, and the intent of 10 CFR 72.104.Response to RAI 10-1In the revised licensing report, the following changes have been made to the Chapter 10operating procedures:a. Section 10.4.1 has been modified such that the step for removing water from theSTC/HI-TRAC annulus (10.4.1.10) is now located immediately before the stepwhere the lid bolting is loosened during the unloading process.b. Section 10.1.2 has been modified to include a step for installation of the BMS onthe HI-TRAC (10.1.2.10). The BMS remains attached to the HI-TRAC and is onlyremoved at the end of the loading campaign or if the pool lid seal needs repair orreplacement.c. Old Step 10.2.3.38, New Step 10.2.3.40 has been modified to include directionsfor monitoring the pressure rise in the STC. Monitoring of the STC pressure willbe performed either remotely or from a low dose area and will not materiallyimpact the estimated occupation dose totals.d. Section 10.5.5 has been modified to provide additional guidance for unloading offuel assemblies suspected of potential damage from an accident condition.e. Section 10.1.3 has been modified to add a step to add water to the HI-TRACneutron shield (10.1.3.5).f. Steps 10.3.1.4 and 10.3.1.8 have been modified to include provisions for controlof access to walkways which may cross over the haul path.g. The types of delays envisioned for requiring water inventory control are asfollows:" Failure of the hydraulic torque system used for the lid bolts that preventsthe STC or HI-TRAC lids hardware from being properly torqued." Crane malfunction which prevents the HI-TRAC lid from being installed inthe HI-TRAC." Leak testing system malfunction that prevents the STC lid seals frombeing leak tested prior to final closure for transfer.Page 77 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONFor each of these delays, the STC will be located in the HI-TRAC. No additionalevaluations or analysis are required for these configurations as they are allbounded by existing analysis of the system provided that the water inventorycontrol is maintained.NRC RAI 10-2Regarding the dose rate measurements for the loaded STC and HI-TRAC, define andinclude the criterion/criteria used to determine when dose rates that exceed expectedvalues may be acceptable and justify waiting to perform dose rate measurements on theHI-TRAC until the currently proposed step. (CSDAB)The currently proposed operations descriptions for the dose rate measurements for theSTC lid and the HI-TRAC side include performance of an evaluation to determine ifhigher dose rates are acceptable and fuel transfer can continue. However, it is not clearwhat criterion or criteria are used to make that determination. Such criteria should bedefined and provided as part of the operations descriptions related to the measurementprocedures. For example, a criterion used for determining the acceptability of the higherdose rates on the transfer cask or storage overpack in the TS Radiation ProtectionProgram for the HI-STORM 100 system is a determination that the as-loaded MPC,considering its contents, the number of casks at the ISFSI, etc. will not cause the limitsof 72.104 to be exceeded. Similarly, an appropriate criterion, or criteria, is needed forthe currently proposed operations. Additionally, the basis for delaying themeasurements on the HI-TRAC side until step 55 of Section 10.2.3, versus performingthe measurements very shortly after the STC is placed in the HI-TRAC (e.g., after step23 or 28), is not clear. If a problem arises that necessitates corrective actions, thevarious operations to prepare the STC and HI-TRAC for transfer will have to be undonewith the current sequence of operations. This seems unnecessary and as well as to notmeet the intent of ALARA. The dose rate measurements on the HI-TRAC side shouldbe performed shortly after the STC is placed in the HI-TRAC.This information is needed to confirm compliance with 10 CFR 20.1101 (b), 10 CFR50.34 and the intent of 10 CFR 72.104.Response to RAI 10-2The steps describing the measurement of the dose rates at the side of the HI-TRAChave been moved up to immediately after the STC has been placed into the HI-TRAC toavoid the potential for additional dose accumulation due to a loading error. The dosemeasurement steps have been revised and refer to the Technical Specification for theexpected dose rate limits and the requirements for determining whether any higher thanexpected dose rates are acceptable. The dose rate limits and the actions required toevaluate the dose readings and their acceptability for continued transfer of fuel betweenunits are defined in Appendix C, Part II, Subsection 5.4 of the Technical Specification.Specifically, Step 18 in Section 10.2.3 has been revised to read:Perform a radiological survey of the STC lid and compare to the expected dose rates asreferenced in Technical Specification Appendix C, Part II, Subsection 5.4, to ensurePage 78 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONthere has not been a fuel mis-load. Dose rate measurements shall be taken at thelocations described in the Technical Specification.Step 19.c has been revised to read:Perform a written evaluation to determine (1) why the surface dose rate limitswere exceeded, and (2) if the higher dose rate values are acceptable, fueltransfer can continue in accordance with controls expressed in TechnicalSpecification Appendix C, Part II, Subsection 5.4.Step 55 has been moved to Steps 23 and 24 and the text has been revised to read:(Step 23) Perform surface dose rate measurements for the HI-TRAC and compare to theexpected dose rates as referenced in Technical Specification Appendix C, Part II,Subsection 5.4. Dose rate measurements shall be taken at the locations described in theTechnical Specification. Compare the measured dose rates with calculated dose ratesfor the design basis fuel to ensure they are less than expected.(Step 24) If dose rates exceed expectations, perform the following:a. Administratively verify that the correct contents were loaded in the correctfuel cell locations.b. Perform a written evaluation to determine (1) why the surface dose ratelimits were exceeded, and (2) if the higher dose rate values areacceptable fuel transfer can continue in accordance with controlsexpressed in Technical Specification Appendix C, Part II, Subsection 5.4.c. If the higher dose rate values are not acceptable, the STC will be returnedto the spent fuel pool and a reload of the STC will be performed.NRC RAI 10-3In Table 10.0.1 of the SAR, "Operational Considerations," add a row for "Crane Hang-upor Loss of Power." This event is discussed on page 10-26, but is not listed in the table.Step 10.5.1.1 .b should start with a requirement to perform radiation surveys to establishstay times for personnel, due to the high dose rates when the STC is suspended fromthe crane. (LPLI-1)Response to RAI 10-3An entry has been entered into Table 10.0.1 for Crane Hang-up or Loss of Power. Step10.5.1.1 .b has been modified to include instructions to perform radiation surveys andestablish radiological controls in the area around the STC.Page 79 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONNRC RAI 10-4In Section 10.5.2 of the SAR, it states that water shall be circulated through the STCdaily to insure that the STC internal cavity is filled. If the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months pressure rise test is inprogress, circulating water or venting will violate the test conditions. Add an exception towater circulation and venting for the pressure rise test. (LPL1 -1)Response to RAI 10-4An exception has been added to Step 1 of Section 10.5.2 of the licensing report toexempt the water circulation requirements during the execution of the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months pressurerise test.Specifically, Step 1 of Section 10.5.2 now includes the following sentence at the end:The above requirements are not applicable during the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months pressure rise test.TECHNICAL SPECIFICATIONS:NRC RAI TS-1Add a TS condition that the restricted area boundary for the transfer operation is aminimum of 20 meters (or the distance used in the evaluations, as modified in responseto this RAI) from the haul path. (CSDAB)The radiation protection evaluation relies upon a set minimum distance to separatemembers of the public on site from the transfer operations to show compliance with 10CFR 20.1301(b). This distance is a significant parameter in the evaluation and shouldbe appropriately controlled. The applicant refers to this distance, or the boundary at thisdistance, as the exclusion (area) boundary (see Section 7.4.6 of Report HI-2094289).This terminology does not seem to be correct for this particular activity. The terminologyshould be made consistent with 10 CFR Part 20, using the terminology defined in thatpart of the regulations for actions performed and controls that are set for purposes ofradiation protection.This information is needed to confirm compliance with 10 CFR 20.1301(b).Response to RAI TS-1A TS condition has been added to proposed TS Section 4.1.4 to state that a restrictedarea boundary will be established at a minimum distance of 20 meters from the haulpath.The minimum distance of 20 meters from the haul path is supported by the analysesdescribed in Section 7.4.7 of the Licensing Report where compliance with 10 CFR20.1301(b) is demonstrated. The term exclusion (area) boundary is no longer used inthis context in the licensing report (HI-2094289) having been replaced by the 10 CFR20.1301 (b) terminology of restricted area.Also note that in response to RAI 7-8 a TS condition has been added to proposed TSSection 4.1.4 that says "During the movement of the STC from the SFP to the HI-TRACPage 80 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONand from the HI-TRAC to the SFP a restricted area boundary shall be established at aminimum distance of 60 meters from the fuel storage building when the STC containsone or more fuel assemblies."NRC RAI TS-2Provide the following clarifications and modifications of the proposed TSs, Appendix C,Parts I and II. (CSDAB)a. Clarify the apparent inconsistency between TS, Appendix C, Part II, 4.1.4.6 andSAR Section 10.2.3.37, modifying the appropriate location to give the correcttolerances on the STC water level.b. Include the STC licensing report (SAR), Chapter 10 as part of the basis inAppendix C, Part I, Section 2.1 and SAR Chapter 8 as part of the basis inAppendix C, Part I, Section 2.2. These sections of the SAR form the basis forthe operations, acceptance tests, and maintenance program directly related tothe STC and the transfer operations.c. Modify LCO 3.1.2.b (Appendix C, Part II) to include an item 4 that duplicates item5 of LCO 3.1.2.a and adds that rod control cluster assemblies (RCCAs) andNSAs cannot be loaded in the configuration of LCO 3.1.2.b. Appropriate Non-Fuel Hardware (NFH) loading restrictions, supported by the licensing reportevaluations, are needed for both loading configurations.d. Clarify the meaning of Note 3 to LCO 3.1.4 (Appendix C, Part II) with regard todefining a given assembly's burnup, especially with respect to the proposed limitsin LCO 3.1.2. It is not clear from this note that assemblies with burnup/exposuregreater than 55 GWD/MTU may not be loaded in the STC (since the noterecalculates the burnup if the assembly had hafnium inserts). The maximumallowable burnup in the TS should be supported by all the appropriate STClicensing report evaluations (e.g., shielding, etc.).e. Add the minimum specifications of the HI-TRAC neutron shielding to the HI-TRAC description in TS, Appendix C, Part I, Section 1.0. The neutron shieldingfeatures are also an important aspect of the HI-TRAC.This information is needed to confirm compliance with 10 CFR Part 50 and the intent of10 CFR 72.44(c), 72.104 and 72.126.Response to RAI TS-2The following clarifications and modifications of the proposed TSs, Appendix C, Parts Iand II are included in the proposed TS accompanying this response:a. The inconsistency between TS, Appendix C, Part II, 4.1.4.6 and SAR Section10.2.3.37, has been resolved. The proposed TS, now LCO 3.1.3, has beenrevised to state that the STC water level shall be 9.0 +0.5/-1.5 inches below thebottom of the STC lid. The associated step in the SAR, now 10.2.3.39, alsostates this water level requirement.b. The STC licensing report (SAR), Chapter 10 has been listed as part of the basisin Appendix C, Part I, Section 2.1 and SAR Chapter 8 has been listed as part ofthe basis in Appendix C, Part I, Section 2.2.Page 81 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONc. LCO 3.1.2.b (Appendix C, Part II) has been modified to include an item 3 thatduplicates the equivalent item of LCO 3.1.2.a (now item 4) and also adds thatRod Control Cluster Assemblies (RCCAs), Neutron Source Assemblies (NSAs),and Hafnium suppressors cannot be loaded in the configuration of LCO 3.1.2.b.Appropriate Non-Fuel Hardware (NFH) loading restrictions, supported by thelicensing report evaluations of Chapters 4 and 7, have been added to proposedLCO 3.1.2, and are included in Table 3.1.2-2 "NON FUEL HARDWARE PostIrradiation Cooling Times and Allowable Average Burnup." Table 3.1.2-2 placesloading restrictions on BPRAs, WABAs, TPDs, RCCAs, Hafnium Suppressors.and NSAs.d. Note 3 to LCO 3.1.5 (Appendix C, Part II) has been clarified with regard todefining a given assembly's burnup, with respect to the proposed limits in LCO3.1.2. This note only applies to the unloading of the fuel from the STC into theUnit 2 spent fuel pool when considering burnup for the application of bumupcredit. The revised Note 3 now says, "For fuel assemblies exposed to Hafniuminserts during irradiation the burnup of the assembly used to classify the fuel forunloading into the IP2 spent fuel pit shall be the burnup prior to the exposure tothe Hafnium insert."Note that the fuel to be loaded into the STC will be limited to the maximumburnups specified in LCO 3.1.2. As discussed in the response to TS-6 theproposed TS limit the radiation source term of the allowable STC contents bylimiting, in part, the maximum allowable burnup.e. The minimum specifications of the HI-TRAC neutron shielding have been addedto the HI-TRAC description in TS, Appendix C, Part I, Section 1.0.NRC RAI TS-3Modify the proposed TS, Appendix C, Part I to: (CSDAB)a. Include recovery from off-normal conditions (such as crane hang-up) in theoperations listed in Section 2.1 andb. Include manual crane operation and crane recovery/repair as part of Section 2.3.The dose rates from the bare STC (analytical) are significant (even for the representativeloading), much more so than has been seen for nearly all spent fuel loading operationsto date. Therefore, staff is particularly concerned about a potential off-normal eventsuch as the hang-up of the crane or a malfunction of the equipment allowing personnelto remain at significant distances from the STC during operations with the STC outsidethe pool and the HI-TRAC. Crane hang-ups are not uncommon, especially when thecranes are loaded with weights approaching their capacity. Thus, operations to recoverfrom a crane hang-up with a loaded STC could provide significant occupationalexposures. Procedures for such scenarios should therefore be developed beforehandand appropriate training provided. Inclusion of manual crane operation in the dry run willassure the licensee can effectively operate the crane manually in a high dose rateenvironment as well as inform the licensee's predictions of potential worker dose in theevent of a crane hang-up or other off-normal event.Page 82 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONThis information is needed to confirm compliance with 10 CFR Part 50 and the intent of10 CFR 72.44(c), 72.104 and 72.126.Response to RAI TS-3In recognition of the importance of recovery from a crane hang up, that could includemanual crane operation, procedures will be developed and training will be provided tosupport such a recovery. In addition, demonstration of manual crane operation will beincluded in the dry run to be performed prior to any fuel transfer operations. Accordingly,the proposed TS, Appendix C, Part I has been modified to:a. Include recovery from all off-normal conditions in the operations listed in Section2.1 Operating Procedures, andb. Include manual crane operation and crane recovery/repair as part of Section 2.3Pre-Operational Testing and Training Exercise.NRC RAI TS-4Provide a TS dose rate limit and measurement requirement for the top lid of the STCand side of the HI-TRAC. (CSDAB)Requirements that establish dose rate limits and the necessary measurements are afeature of the TS associated with 10 CFR Part 72 dry storage loading operations(including the HI-STORM 100), which are similar to the loading operations performed forthe proposed wet transfer system. TS dose rate limits and measurements provideassurance of correct contents loading, ensure operations and ALARA planning is stilladequate/appropriate for a given loading operation, and ensure that conditions outsidethose assumed for the design and operations are identified and properly handled toassure protection of personnel and members of the public.Considering that the dose rates for even the proposed contents are very substantial andthat conditions of the kind noted herein could make them even more substantial, a TSshould be established that provides dose rate limits for the STC lid and the HI-TRACside, an appropriate measurement scheme for each limit, and the appropriate correctiveactions for instances where the limits are exceeded. The limits should be derived fromthe analysis for the representative loading and should capture areas of the STC and HI-TRAC of significance to occupational and public dose. Given the similarities of theproposed system to dry storage transfer systems, staff anticipates that the limits andmeasurements will be established in a similar fashion as for dry storage transfersystems, with appropriate consideration for differences versus those systems and theapplicant's analyses serving as the basis.If it is proposed that dose rates that are higher than the proposed limits may beevaluated and considered acceptable under appropriate circumstances, the process andcriteria for finding higher dose rates acceptable and allowing continuance of operationsshould be included in the TS and appropriately justified. For example, for dry storage,higher dose rates trigger a verification of correct loading and, if the loading is correct, adetermination of whether or not 10 CFR 72.104 limits can be met for the loaded MPC.Since this is a wet transfer between pools, a criterion (such as ensuring compliance with10 CFR 20.1301(b)) in addition to 72.104 limit compliance, considering the number oftransfers to be performed in a given year, may be appropriate.Page 83 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONThis information is needed to confirm compliance with 10 CFR 50.34, 10 CFR 20.1101and 20.1301(a) and (b), and the intent of 10 CFR 72.104.Response to RAI TS-4TS dose rate limits and measurement requirements for the top lid of the STC and side ofthe HI-TRAC have been developed and are included in proposed TS 5.4 RadiationProtection Program. The limits have been derived from the analysis for therepresentative loading with the highest expected dose rates and capture areas of theSTC and HI-TRAC of significance to occupational and public dose. These limits andmeasurements have been established in a similar fashion as for dry storage systems.It is proposed that dose rates that are higher than the proposed limits may be evaluatedand considered acceptable under appropriate circumstances. The process and criteriafor finding higher dose rates acceptable and allowing continuance of operations areincluded in proposed TS 5.4 and are appropriately justified. Higher dose rates trigger averification of correct loading (TS 5.4.4.a.) and, if the loading is correct, a determinationof whether or not the transfer can proceed without exceeding the dose limits of 10 CFR72.104 or 10 CFR 20.1301 (TS 5.4.4.b.).The fuel transfer dose is listed separately in the site boundary dose report prepared inaccordance with 10 CFR 72.104. The dose is considered part of the plant's operationalcontribution to the site boundary dose. Currently the site boundary dose report assumesthat there are 16 inter-unit fuel transfers per year with a bounding dose for each transfer.If dose rate measurements taken in accordance with Section 5.4 are exceeded anevaluation is expected which will consider the assumptions already in the site boundarydose report to ensure that the yearly limit will not be exceeded. Similarly, the dose ratelimits in 10 CFR 20.1301 will also be considered and an evaluation performed to ensurethat occupational and public doses will not be exceeded.NRC RAI TS-5Modify the fuel specifications in proposed TS, Appendix C, Part II, Table 4.1.1-1 toaccurately reflect the contents (as evaluated in the amendment application) that will beloaded into the STC, including the following changes. (CSDAB)a. Include the fuel cladding material. The shielding evaluation only supportszirconium-based cladding; thus, this should be the listed cladding.b. Change the fuel rod clad I.D. to be a maximum value. It is currently shown as aminimum value, which appears to be incorrect. A minimum cladding thicknessshould be established, which only can be done with the clad I.D. set at amaximum and the clad O.D. set as a minimum.c. The correct maximum active fuel length should be given. The shieldingevaluation, for example, indicates the maximum active fuel length is 144 inches.d. The correct fuel assembly maximum length should be given. The current length(176.8 inches) given in the table does not physically fit in the STC, which thelicensing drawings indicate has a cavity length of only 168 15/16 inches. Further,the inclusion of NFH extends the needed STC cavity length to be able to containthe proposed contents.Page 84 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONe. If the active length and assembly length are accurate, then the STC designshould be revised and the evaluations modified accordingly to support the newdesign and the proposed contents.This information is needed to confirm compliance with 10 CFR 50.34 and the intent of 10CFR 72.104.Response to RAI TS-5The fuel specifications in proposed TS, Appendix C, Part II, Table 4.1.1-1 have beenmodified to accurately reflect the contents (as evaluated in the amendment application)that will be loaded into the STC, including the following changes:a. The fuel cladding material.b. The fuel rod clad I.D. to be a maximum value.c. The correct maximum active fuel length is given. The fuel along with NFH willphysically fit in the STC. Fuel spacers are typically used in MPCs for dry storageand transportation to keep the active length of the fuel aligned with the neutronabsorber panels in the basket during horizontal handling and eventual transportof the MPC. In the horizontal position it may be possible for the fuel to move uptowards the top of the MPC and outside of the envelope of the neutron absorberpanels. Fuel spacers are not required for the STC since there are no instanceswhere the STC will be in horizontal position.d. The correct fuel assembly maximum length is given.e. No design changes and no additional evaluations are necessary based onresponses to c. and d. above.NRC RAI TS-6Propose TS limits on maximum burnup, minimum enrichment and minimumdecay/cooling times to limit the radiation source term of the contents and providesupporting quantitative evaluations to justify the proposed limits. Otherwise, provideadditional justification, including appropriate quantitative evaluations, that the currentlyproposed TS with respect to allowable STC contents are sufficient to limit the radiationsource of those contents. (CSDAB)The currently proposed operation is similar to loading operations for dry storagesystems. TS for these systems, including the HI-STORM 100, define the allowablecontents in terms of the maximum burnup, minimum enrichment and minimum decaytime of the assemblies. That being the case, the applicant has proposed a very differentapproach to limit the contents. The applicant has provided some justification in responseto the first round RAI question 10-5. However, no quantitative evaluation was providedto support the statements made in that response. While the applicant has included asource term that significantly exceeds the decay heat limits for comparison as part of thelicensing report, this does not indicate the variation in dose rates that will exist betweenassemblies of different burnup, decay time, and minimum enrichment that have thesame decay heat. Adequate justification would include evaluations that show how thedose rates vary for an adequate variety of assembly burnup, enrichment and decay timecombinations that result in the same decay heats for the different STC basket regions.Page 85 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONGiven that radiation source and decay heat do not correlate on a one-to-one basis andthe significance of the STC dose rates for the 'representative' loading, it is important toensure the TS properly define the allowable radiation source in the STC. A decay heatlimit alone would allow for radiation sources of varying strengths that could result in doserates that may be different enough to necessitate non-trivial changes to operationprocedures and controls for the purposes of occupational and public radiation protectionand significantly impact off-normal conditions and accident conditions exposures as well.Additionally, it is not clear how using a decay heat limit alone would be practicallyimplemented, including independent verifications, considering that decay heat is acalculated value that is derived from an assembly's enrichment, cooling/decay time andburnup.This information is needed to confirm compliance with 10 CFR 20.1101(b), 10 CFR50.90 and 50.34a(c) and the intent of 10 CFR 72.104, 72.106 and 72.44(c)(1).Response to RAI TS-6TS limits on maximum burnup, minimum enrichment and minimum decay/cooling timesare proposed to limit the radiation source term of the allowable STC contents.The analyzed loading configurations are provided in the Table TS 1 below (Table 7.1.1of the licensing report), and bound the majority of the current Unit 3 spent fuel poolinventory. The remaining assemblies will only be transferred when their cooling timesmeet the requirements of Table TS 1. As determined in Section 7.4 of the licensingreport, loading pattern 3 or 4 results in the bounding dose rates depending on the doselocations and whether the STC is outside or inside the HI-TRAC.The source terms were applied in a regionalized loading scheme to the 12 fuel assemblylocations available in the STC. The regionalized loading pattern was utilized to facilitatethe transfer of hotter fuel in the spent nuclear fuel storage pool by taking advantage ofself-shielding effects. As can be seen in Table TS 1 the source terms with the highercooling times are assigned to the eight outer fuel assembly locations in the STC, whilethe source terms with the lower cooling times are assigned to the four inner fuelassembly locations.Therefore, it is proposed to revise LCO 3.1.2 to include the limits on maximum burnup,minimum enrichment and minimum decay/cooling times specified in Table TS 1. ThisTable has been added to LCO 3.1.2 as Table 3.1.2-3.Page 86 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONTable TS 1Allowable STC Loading ConfigurationsConfiguration Cells 1, 2, 3, 4(a)(b) Cells 5, 6, 7, 8, 9,10,11, 12(a)(b)Burnup < 55,000 MWD/MTU Burnup < 40,000 MWD/MTU1 Cooling time > 10 years Cooling time > 25 yearsInitial Enrichment > 3.4 wt% U-235 Initial Enrichment > 2.3 wt% U-235Burnup < 45,000 MWD/MTU Burnup < 45,000 MWD/MTU2 Cooling time > 10 years Cooling time > 20 yearsInitial Enrichment > 3.2 wt% U-235 Initial Enrichment > 3.2 wt% U-235Burnup < 55,000 MWD/MTU Burnup 5 45,000 MWD/MTU3 Cooling time > 10 years Cooling time > 20 yearsInitial Enrichment > 3.4 wt% U-235 Initial Enrichment > 3.2 wt% U-235Burnup 5 45,000 MWD/MTU Burnup 5 40,000 MWD/MTU4 Cooling time > 10 years Cooling time a 12 yearsInitial Enrichment > 3.6 wt% U-235 Initial Enrichment > 3.2 wt% U-235Burnup 5 45,000 MWD/MTU Burnup 5 40,000 MWD/MTU5 Cooling time a 14 years Cooling time > 12 yearsInitial Enrichment > 3.4 wt% U-235 Initial Enrichment a 3.2 wt% U-235(a) Natural or enriched uranium blankets are not considered in determining the fuelassembly enrichment for comparison to the minimum allowed initial enrichment.(b) Rounding to one decimal place to determine initial enrichment is permitted.NRC RAI TS-7Justify the lack of a TS surface contamination LCO for the STC. (CSDAB)Operations include the immersion of the STC in the spent fuel pool (SFP). While theSTC is washed down prior to and as it is removed from the SFP, this is to minimizecontamination. This is the same kind of practice taken with the transfer cask for spentfuel loading operations for dry storage and there is normally a TS LCO limiting surfacecontamination (see TS 3.2.2 for the HI-STORM 100 system). The bases for that LCOstate the LCO "allows dry fuel storage activities to proceed without additional radiologicalcontrols to prevent the spread of contamination and reduces personnel dose due to thespread of loose contamination or airborne contamination." While the HI-TRAC for thePage 87 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONcurrently proposed operations does not enter the SFP, the STC acts as a transfer caskfor the loading and unloading operations in its use out of the HI-TRAC. Given thisconsideration and the leak rates allowed for the HI-TRAC, a contamination LCO may beappropriate.This information is needed to confirm compliance with 10 CFR 20.1101(b), 10 CFR50.34, and the intent of 10 CFR 72.44(c)(1), 72.104, and 72.126(a).Response to RAI TS-7In the HI-STORM 100 System LCO 3.2.2 addresses transfer cask surfacecontamination. LCO 3.2.2 notes that the LCO is not applicable to the transfer cask if itremains in the fuel storage building. In this situation the outside surface of the transfercask would not be exposed to the environment. An equivalent to LCO 3.2.2, though, isnot applicable to the STC as the STC is not a final dry fuel storage canister, nor is itsubject to weathering. By virtue of its intended use, the STC outside surface would notbe exposed to the environment even after immersion in either spent fuel pool. When theSTC leaves a building it is confined within the HI-TRAC.As noted in the RAI, the basis for the HI-STORM 100 System LCO 3.2.2 is that the LCO"allows dry fuel storage activities to proceed without additional radiological controls to...prevent the spread of contamination and reduces personnel dose due to the spread ofcontamination or airborne contamination." As also noted in the RAI, the STC iswashed down as it is removed from the SFP, to minimize contamination. Entergy has notproposed an LCO equivalent to the HI-STORM 100 System LCO 3.2.2 due to ALARAconcerns that arise from performing the decontamination activity together with the factthat the prevention of the spread of contamination and the reduction in personnel doseare concerns applicable to the immediate transfer activity and not to long term storage.Radiological controls will remain in effect during the transfer of fuel. Large area swipeswill be taken of the STC and HI-TRAC surfaces in accordance with proceduralradiological controls specifically developed for this activity. The intent of these actions,monitoring and controls is to reduce and detect loose contamination and particles andallow timely and appropriate personnel protection action. The level of effort of thesecontamination and particle surveys will be in accordance with operational ALARAprinciples.Contamination in the form of particles could be transferred from the surface of the STCto the water in the annulus space inside the HI-TRAC; however, the HI-TRAC pressureboundary will be tested to ensure that it is water tight to prevent the loss of water fromthe annular region during fuel transfer operations. This testing is described in Section 8.4of the licensing report. Following completion of an inter-unit fuel transfer campaignEntergy expects to have the HI-TRAC restored to such a condition to support all drytransfer operations and conditions of that mode of operation.NRC RAI TS-8Revise the proposed TS 3.1.3 to ensure a correct measurement of the pressure rise inthe STC after it is loaded with spent fuel. (LPLI -1)Measurement of the pressure change in the STC cavity after the STC is loaded is usedto verify that the decay heat load is within design parameters. The staff's understandingPage 88 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONof the pressure rise monitored by proposed LCO 3.1.3 is that the pressure rise is thedifference between the lowest observed absolute pressure in the STC and the absolutepressure in the STC at the end of the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. That description of the pressurerise should be added to TS Bases 3.1.3. Also, since the staff expects the pressure inthe STC to initially decrease after the water level is established, SR 3.1.3.1 should berevised to say, "Once upon establishing required water level AND hourly thereafter."Taking at least 25 data points over the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period will allow more accurate checkingof the pressure change. Also, SR 3.1.3.2 should be added to specify the pressureinstrumentation to be used. Wording similar to the following is needed: 'Verify that twochannels of pressure instrumentation with a range of at least 1 psia to 75 psia, andcalibrated to within 2% accuracy within the past 12 months, are installed on the STC."The Frequency of SR 3.1.3.2 could be "During performance of surveillance 3.1.3.1."This information is required for compliance with 10 CFR 50, Appendix A, GDC 61.Response to RAI TS-8The proposed TS (now LCO 3.1.4) has been revised to ensure a correct measurementof the pressure rise in the STC after it is loaded with spent fuel. As discussed in theresponses to RAIs 1-1, 5-5, and 5-6, it is proposed to detect a significant fuel misload bymeasuring the rate of pressure rise over a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. LCO 3.1.4 has been revisedto say, "The pressure rise in the STC cavity shall be < 0.2 psi/hr averaged over a rolling4 hour period."A rate of STC pressure rise above 0.2 psi/hr during any rolling 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period would be anindication that the design basis of the STC is not being met. This indication could bedue to one or more of the following conditions:a. STC water level not within limit,b. one or more fuel assemblies misloads,c. presence of air in the STC cavity.Should the acceptance criterion of 0.2 psi/hr not be met the STC would bedepressurized and actions taken to verify: the STC water level, that a fuel misload hadnot occurred, and that air had been effectively precluded from entering the STC. Oncethe cause of the non conforming condition has been identified and corrective actionstaken, SR 3.1.4.1 would need to be re-performed satisfactorily prior to STC transferoperations.Measurement of the rate of pressure rise in the STC cavity after the STC is loaded isused to verify that fuel transfer can proceed within the established thermal design basis.It is proposed to take 25 data points over the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. SR 3.1.4.1 has beenrevised to say, "Verify by direct measurement that the rate of STC cavity pressure rise iswithin limit." A Note has also been added to the SR that says, "Pressure measurementsshall be taken once upon establishing required water level AND hourly thereafter for 24hours. Pressure may initially drop during pressure stabilization."SR 3.1.4.2 has been added to specify the pressure instrumentation to be used and thatan ASME Code compliant pressure relief valve or rupture disc must be installed duringthe test. The proposed SR says, "Verify that an ASME code compliant pressure reliefvalve or rupture disc and two channels of pressure instrumentation with a range of atPage 89 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONleast 0.1 psia to 15 psia and calibrated to within 1% accuracy within the past 12 monthsare installed on the STC." The frequency of SR 3.1.4.2 is "During performance of SR3.1.4.1."Note that two sets of pressure instrumentation are required for the operation of STC asshown in Table 10.1.1. The use of each is described in Chapter 10. The pressure risegauges are as specified above and are used during the LCO Surveillance. The other setof pressure gauges also referred to as the "coarse" gauges, are used along with theASME code compliant relief valve or rupture disk during leak testing and otheroperations involving the STC to ensure no over-pressurization of the STC occurs.NRC RAI TS-9Throughout the TS, "non fuel hardware" is a defined term and should therefore becapitalized. (LPLI-1)Response to RAI TS-9It is recognized that "non fuel hardware" is a defined term and the proposed TS havebeen revised accordingly.NRC RAI TS-10SR 3.1.1.1 says to "Verify the STC boron concentration is within limit using twoindependent measurements." Please explain the independent measurements to beused. If they are not truly independent (such as titration and neutron absorption), then itmay be better to describe them as "two separate measurements." (LPLI-1)This information is required for compliance with 10 CFR Part 50, Appendix A, GDC 61.Response to RAI TS-10The STC boron concentration will be verified using two separate measurements andproposed SR 3.1.1.1 has been revised to say, "Verify the boron concentration is withinlimit using two separate measurements." Note that LCO 3.1.1 has been revised toinclude the boron concentration of the water in the spent fuel pit in addition to the STC.NRC RAI TS-11TS 4.1.4.3 says "LOADING OPERATIONS shall only be conducted when the IP3 spentfuel pit contains irradiated fuel only." Consider revising this to say "LOADINGOPERATIONS shall only be conducted when the IP3 spent fuel pit contains nounirradiated fuel." This is more precise, since there typically are items other thanirradiated fuel in the spent fuel pit. (LPL1-1)This information is required for compliance with 10 CFR Part 50, Appendix A, GDC 61.Page 90 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONResponse to RAI TS-11The spent fuel pit typically does contain items other than irradiated fuel, therefore,proposed TS 4.1.4.3 has been revised to say, "LOADING OPERATIONS shall only beconducted when the IP3 spent fuel pit contains no unirradiated fuel assemblies."NRC RAI TS-12TS 4.1.4.8 says "TRANSFER OPERATIONS shall only be conducted when the HI-TRACwater level is within +0/-1 inch of the top of the STC lid and the water level has beenindependently verified." Consider revising this to say "TRANSFER OPERATIONS shallonly be conducted when the HI-TRAC water level is within +0/-1 inch of the top of theSTC lid prior to installing the HI-TRAC lid and the water level has been independentlyverified." It is obvious that as the water heats up in the HI-TRAC it will expand and will nolonger be within the specified range. (LPLI-1)This information is required for compliance with 10 CFR Part 50, Appendix A, GDC 61.Response to RAI TS-12The HI-TRAC water level is established and independently verified prior to installing theHI-TRAC lid, therefore, the proposed TS, which is now 4.1.4.9, has been revised to say,"Prior to installing the HI-TRAC lid the HI-TRAC water level shall be verified by twoseparate inspections to be within +0/-1 inch of the top of the STC lid."NRC RAI TS-13TS 5.2(iii) says "Four coupons will be tested at the end of each inter-unit fuel transfercampaign." Since the duration of a campaign is not defined, please propose a timeframe instead, such as every 2 years. (LPL1-1)This information is required for compliance with 10 CFR Part 50, Appendix A, GDC 61.Response to RAI TS-13Proposed TS 5.2(iii) has been revised to say, "Four coupons shall be tested at the end ofeach inter-unit fuel transfer campaign. A campaign shall not last longer than two years.The coupons shall be measured and weighed and the results compared with the pre-characterization testing data. The results shall be documented and retained."NRC RAI TS-14The Note in SR 3.1.1. says that the surveillance is only required to be performed if theSTC is submerged in water. Since the STC is submerged in water while it is in the HI-TRAC, the boron concentration would have to be checked every 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . If the STChad to be left in the HI-TRAC for an extended period of time, for example, due to anequipment malfunction, it will be very difficult to get a sample for boron analysis.Recommend revising the wording to something similar to "This surveillance is onlyrequired to be performed if the STC is submerged in water with the STC lid notfastened...". (LPL1-1)Page 91 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONThis information is required for compliance with 10 CFR Part 50, Appendix A, GDC 61.Response to RAI TS-14Originally proposed SR 3.1.1 is modified by a Note indicating that the surveillance is onlyrequired to be performed if the STC is submerged in water or if water is to be added to,or recirculated through, the STC. These are the only times when a change in STC boronconcentration could potentially occur. In order to preserve the assumptions of thecriticality analysis, the Note further requires that water added to, or recirculated through,the STC must meet the boron concentration requirements of LCO 3.1.1. This Note doesnot apply to the addition of steam to the STC as discussed in the licensing report.In recognition of the fact that the STC could be submerged in water in the HI-TRAC foran extended period of time without potential for a change in boron concentration, theNote in proposed SR 3.1.1 has been revised to say, "This surveillance is only required tobe performed if the STC is submerged in water in the spent fuel pool or if water is addedto, or recirculated through, the STC when the STC is in the HI-TRAC. Any added watermust meet the boron concentration requirement of LCO 3.1.1."NRC RAI TS-15LCO 3.1.1 would allow the licensee to sample the boron in the STC to verify at least2000 ppm, and then lower the STC into the IP3 spent fuel pit with the water in the spentfuel pit at 1000 ppm boron. This allows the possibility of diluting the STC boron. Eitherrevise Appendix A, LCO 3.7.15, to require 2000 ppm in the IP3 spent fuel pit wheneverfuel assemblies are in it, or add an LCO to Appendix C to verify 2000 ppm in the IP3spent fuel pit prior to placing the STC in the spent fuel pit. Revise SAR section 1.4, andother sections as necessary, to reflect this change. (LPL1-1)This information is required for compliance with 10 CFR Part 50, Appendix A, GDC 61.Response to RAI TS-15The intent of LCO 3.1.1 and the associated SR was not to allow sampling of the STCboron concentration prior to lowering the STC into the IP3 spent fuel pit. Rather, theintent was to demonstrate that the STC boron concentration, when the STC issubmerged in the IP3 spent fuel pit, is greater than or equal to 2000 ppm, prior toloading fuel into the STC.In order to clarify this intent the Note to SR 3.1.1.1 has been revised to read, "Thissurveillance is only required to be performed if the STC is submerged in water in thespent fuel pool or if water is added to, or recirculated through, the STC when the STC isin the HI-TRAC. Any added water must meet the boron concentration requirement ofLCO 3.1.1." The specified frequency of SR 3.1.1.1 is "Once, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior toentering the applicability of this LCO and once per 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months thereafter." This frequency,together with the LCO, ensures that the boron concentration of the water in the STC inthe spent fuel pool is verified to be greater than or equal to 2000 ppm, prior to loadingfuel into the STC.In order to ensure that the boron concentration of the water in the spent fuel pit is notreduced below that assumed in the STC criticality analysis LCO 3.1.1 has been revisedPage 92 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONto say, "The boron concentration of the water in the Spent Fuel Pit and the STC shall be 2000 ppm." In addition, it is proposed to revise Unit 3 LCO 3.7.15, "Spent Fuel PitBoron Concentration" by adding the following Note: "During inter-unit transfer of fuel thespent fuel pit boron concentration must also meet Appendix C LCO 3.1.1, "BoronConcentration.""Sections 10.2 and 10.4 Inter-Unit Transfer Operations of the licensing report has beenrevised to clarify when and where boron concentration measurements are taken.NRC RAI TS-16TS 4.1.2.1 is insufficient to specify the criticality controls, as it does not specify thelocation or size of the Metamic panels. One method would be to incorporate a drawingby reference which shows that information. Also, TS 4.1.2.1 .g refers to B-10 loading inthe B4C as greater than or equal to 18.4 %. This can be misinterpreted, as 18.4 weightpercent of the B4C is not B-10. Please revise it to specify the B-10 density in the B4C interms of an areal density (grams per square centimeter). (LPLI-1)This information is required for compliance with 10 CFR Part 50, Appendix A, GDC 61.Response to RAI TS-16The parameter most important for the criticality control function of the neutron absorberis the so-called B-10 areal density, i.e. the amount of B-10 per unit area of the absorberpanel (usually specified as gm B-1 0/cm2). While this parameter can be measured in thefinal product (via neutron attenuation testing), it is not a direct input into themanufacturing process. However, the value is the mathematical product of three inputand process parameters, namely the B4C weight percent of the material, the percent B-10 in the Boron in the B4C, and the thickness of the panel, together with an appropriateproportionality constant (See Section 8.2 of the licensing report).To provide a robust and conservative acceptance criteria approach, each of these threeparameters is controlled independently in the manufacturing process and eachparameter must independently meet a specified minimum required value.This approach essentially guarantees that the panels exceed the required areal density.Since the approach uses a worst-case combination of the minimum value for eachparameter, no statistical evaluation or criteria is required. The specific requirements aretherefore:* All lots of B4C will contain boron with an isotopic B-10 content of at least 18.4%." The B4C content in METAMIC shall be greater than or equal to 31.5 and less thanor equal to 33.0 weight percent. (on drawing)" The Metamic panel thickness must be no less than the minimum thickness specified,0.102 in. (on drawing)These three measurable parameters can be verified by the documentation packagesand QA records for the panels used in the STC. This assures that the minimum B-10areal density of 0.031 g/cm2 is achieved. The above three parameters are given in theSTC basket drawing 6015 along with the size and location of the panels on the basket.Proposed TS 4.1.2.1.i has been added to reference drawing 6015 for the size andlocation of the panels and says "The size and location of the neutron absorber panelsPage 93 of 94 HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONshall be in accordance with drawing 6015 of the Licensing Report (Holtec InternationalReport HI-2094289)."Proposed TS 4.1.2.1.g has been modified to state more clearly that the B4C will containboron with an isotopic B-1 0 content of at least 18.4% and now says "The B4C in theMetamic neutron absorber will contain boron with an isotopic B-10 content of at least18.4%".NRC RAI TS-17One of the operating conditions of the STC is the possibility of leaving some fuelassembly locations open in order to allow the loading of fuel which does not meet theminimum burnup requirements in the remaining locations. This is, therefore, an initialcondition of a design-basis accident (fuel misload). This should be controlled by a TS.Please revise the Note for LCO 3.1.2 to say that if one or more Type 1 fuel assembliesare in the STC, cells 1, 2, 3, and 4 must be empty, with cell blockers installed thatprevent inserting fuel assemblies. Also propose a new SR 3.1.2.2 that verifies by visualinspection that cell blockers are installed on cells 1, 2, 3, and 4 prior to placing a Type 1fuel assembly in the STC. Add the description of the cell blockers to the SAR. (LPL1 -1)This information is required for compliance with 10 CFR Part 50, Appendix A, GDC 61.Response to RAI TS-17In order to allow the loading of fuel assemblies that do not meet minimum burnuprequirements (Type 1 assemblies) some STC fuel assembly locations are left open asspecified in proposed LCO 3.1.2. Therefore, in order to preclude the physical placementof fuel assemblies in these open locations, a cell blocker will be installed as necessary.The Note for LCO 3.1.2 has been revised to say, "If one or more Type 1 fuel assembliesare in the STC, cells 1, 2, 3, and 4 must be empty, with a cell blocker installed thatprevents inserting fuel assemblies."A new SR 3.1.2.2 is also proposed that says, "Verify by visual inspection that a cellblocker which prevents inserting fuel assemblies into cells 1, 2, 3, and 4 of the STC isinstalled." The frequency of the proposed SR is "Prior to placing a Type 1 fuel assemblyin the STC."A figure depicting a typical cell blocker has been added to Chapter 10 of the LicensingReport, Figure 10.2.Page 94 of 94

@@ Line 13: / Line 13: @@
 | document type = Request for Additional Information (RAI)
 | page count = 94
-| project = TAC:ME1671, TAC:ME1672
+| project = TAC:ME1672, TAC:ME1671
 | stage = RAI
 }}
 =Text=
-{
+{{#Wiki_filter:HOLTEC INTERNATIONAL NON-PROPRIETARY INFORMATIONREQUEST FOR ADDITIONAL INFORMATIONREGARDING SPENT FUEL
 }}

v t e Letter Sequence Request
TAC:ME1672, Control Room Habitability (Open) TAC:ME1671, Control Room Habitability (Open)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance Supplement, Supplement, Supplement, Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance Meeting, Meeting, Meeting Questions 20 April 2010 27 May 2010 16 November 2010 16 March 2011 28 September 2011 9 November 2011 Answers 28 October 2010 23 August 2011 28 October 2011 11 January 2012 5 October 2010 28 July 2011 28 October 2011 15 December 2011 2 March 2012 Results Approval Other: ML092090289, ML092330159, ML120740550, ML121230011, NL-10-122, Enclosure 1 to NL-10-122, Holtec International Licensing Report HI-2094289, Rev. 3, Licensing Report for Transfer of IP-3 Fuel to IP-2, NL-11-100, HI-2094289, Revision 4, Licensing Report on the Inter-Unit Transfer of Spent Nuclear Fuel at the Indian Point Energy Center, Enclosure 2 to NL-11-100
MONTHYEAR2010-04-20 [Table View]

ML11243A175: Difference between revisions

Revision as of 15:57, 5 April 2018

Text

Reference:

Reference:

Reference:

Reference:

Reference:

Navigation menu

ML11243A175: Difference between revisions

Revision as of 15:57, 5 April 2018

Text

Reference:

Reference:

Reference:

Reference:

Reference:

Navigation menu

Search