ML14077A303

ANP-3289NP, Rev. 0, Responses to RAI from Snpb on MNGP Transition to Areva Fuel.

ML14077A303
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Issue date:	02/28/2014
From:	AREVA
To:	Office of Nuclear Reactor Regulation
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L-MT-14-022, TAC MF2479 ANP-3289NP, Rev 0
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{{#Wiki_filter:L-MT-14-022ENCLOSURE 2AREVA REPORT NO. ANP-3289NP, REVISION 0RESPONSES TO RAI FROM SNPB ON MNGP TRANSITION TO AREVA FUELNON-PROPRIETARY36 pages follow ANP-3289NPRevision 0Responses to RAI from SNPB onMNGP Transition to AREVA FuelFebruary 2014AAREVAAREVA Inc. AREVA Inc.ANP-3289NPRevision 0Responses to RAI from SNPB onMNGP Transition to AREVA Fuel AREVA Inc.ANP-3289NPRevision 0Copyright © 2014AREVA Inc.All Rights Reserved Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page iNature of ChangesItem1.PageAllDescription and JustificationThis is the initial issueAREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page iiContents1.02.03.0In tro d u c tio n ............................................................................................................................. 1-1R A Is a nd R e sp o nse s .............................................................................................................. 2-1R e fe re n c e s ............................................................................................................................. 3 -1TablesTable 1 Monticello Thermal-Hydraulic Results at Rated Conditions (100%P / 105%F) forT ransition to A T R IU M 1OX M Fuel ........................................................................................... 2-9Table 2 Monticello Thermal-Hydraulic Results at Rated Conditions (100%P / 80%F) forT ransition to A T R IU M 1OX M Fuel ......................................................................................... 2-10Table 3 Monticello Thermal-Hydraulic Results at Off-Rated Conditions (82.5%P /57.4%F) for Transition to ATRIUM 1OXM Fuel ...................................................................... 2-11Table 4 MNGP ATRIUM 1OXM Fuel Assembly .................................................................................. 2-21FiguresFigure 1 -Calculated Cladding Collapse Margin to Fuel Column Axial Gap Limit ................................... 2-3Figure 2 -MCPR Penalty Model vs. Test Data ..................................................................................... 2-18AREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 1-11.0 IntroductionIn Reference 1, Northern States Power Company -a Minnesota corporation, doing business as XcelEnergy, requested an amendment to the operating license and facility Technical Specifications for theMonticello Nuclear Generating Plant (MNGP). The amendment, if approved, would allow for a transition tothe AREVA ATRIUM 1OXM fuel design. The amendment would also allow the implementation of AREVAsafety analysis methods.The U.S. Nuclear Regulatory Commission (NRC) staff in the Nuclear Performance and Code Reviewbranch (SNPB) is reviewing the safety analyses for anticipated operational occurrences (AOOs), designbasis accidents (DBAs), and special events. The SNPB staff has determined that additional information isrequired to complete its review (Reference 2). The Requests for Additional Information (RAI) and theAREVA responses are attached.These responses are provided so Xcel Energy can provide a complete set of responses to the NRC bycombining the AREVA responses with the responses being prepared by Xcel Energy.AREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page 2-12.0 RAIs and ResponsesSNPB RAI-1: ANP-3221P, Section 3.2.2Please provide a detailed description of the statistical method used in the analysis for cladding creepcollapse that yielded the best-estimate results for the creep collapse of the cladding.AREVA ResponseThe creep collapse criterion is evaluated by calculatihg the formation of an axial gap in the fuel columndue to fuel densification for comparison to the design limit on gap size. The design criterion is as follows.Clad creep collapse shall be prevented. [The RODEX4 application methodology is a statistical uncertainty propagation method that uses a MonteCarlo random sampling of relevant input parameters to evaluate the propagation of the uncertainties tothe design analysis results. []The uncertainties used in the analysis are categorized as [IAREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page 2-2IAREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page 2-3IFigure 1 -Calculated Cladding Collapse Margin toFuel Column Axial Gap LimitThe above description of the methodology and criteria are consistent with the approved RODEX4 topicalreport, BAW-10247P (Reference 3). More specifically, details on the methodology can be found in an RAIresponse during the review of RODEX4, Response 18 on page 84 of BAW-10247Q4P that is containedwithin the approved topical report.SNPB RAI-2: ANP-3221P, Section 3.2.3In the methodology for analysis of overheating of fuel pellets, it is stated that linear heat generation rate(LHGR) margins are provided along with LHGR uncertainties due to channel bow input to the statisticalanalysis.Please provide details of how the channel bow uncertainties are developed and how they areincorporated in to the statistical analysis.Note: This RAI is similar to the one staff made for the review of ANP-3159P which was part of the BrownsFerry units' fuel transition to ATRIUM IOXM fuel design.AREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 2-4AREVA ResponseThe uncertainty in the calculated channel bow leads to an associated uncertainty in the fuel rod powerlevel. This uncertainty in power is taken into account as part of the RODEX4 statistical applicationmethodology. A series of steps are carried out to assess the effect of channel bow and its associatedmodel uncertainty on the fuel rod thermal-mechanical behavior by accounting for channel bow in thegeneration of the fuel rod power histories.[]AREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 2-5[]The above description is consistent with the methodology described in Reference 3. Additionalinformation can be found in the third round of RAI responses, BAW-10247Q3(P), that is contained in theapproved RODEX4 topical report.The RODEX4 results presented in ANP-3221P for the Monticello Cycle 28 ATRIUM 1OXM fuel include theadjustments as described above to account for power uncertainties from channel bow. The method isidentical to that used for the RODEX4 calculations in support of the Browns Ferry LAR and the Brunswickreload licensing calculations.SNPB RAI-3: ANP-3221P, Section 3.2.7Please provide details of how RODEX4 treats heat transfer coefficient to account for the presence of crudat normal, low level and abnormal level, and explain how this is applied to crud measurements at MNGP.AREVA ResponseThe RODEX4 code includes a provision to input a crud thickness layer and the associated thermalconductivity of the crud. This feature was briefly described in one of the RAI responses in the RODEX4topical report (BAW-10247Q4P, Response 15, p77, Reference 3). If input, the crud layer is modeled as anadditional thermal resistance term between the coolant film and the corrosion layer. The input to RODEX4for a crud layer depends on whether the plant is considered to have low, normal levels of crud or if higherlevels of crud are indicated.The RODEX4 model for calculating corrosion is described as an oxidation model and it calculates acorrosion layer using a relation based on the kinetics of oxidation. However, it would be more accurate tocharacterize the model as predicting "liftoff' because the model is benchmarked to liftoff measurementdata and not solely corrosion data. The term liftoff refers to the separation or liftoff of the eddy currentmeasurement probe from the metallic surface of the fuel rod due to the presence of the insulatingcorrosion and crud layers. The liftoff measurement method cannot discern between oxide and crud. Visualexamination of the fuel rod surfaces, more recent hot cell data, and crud scrape measurements confirmthe presence of the thin, tenacious crud layers.BAW-10247PA SER restriction 5 on crud says:"RODEX4 has no crud deposition model. Due to the potential impact of crud formation on heattransfer, fuel temperature, and related calculations, RODEX4 calculations must account for adesign basis crud thickness. The level of deposited crud on the fuel rod surface should be basedupon an upper bound of expected crud and may be based on plant-specific history. Specificanalyses would be required if an abnormal crud or corrosion layer (beyond the design basis) isobserved at any given plant. For the purpose of this evaluation, an abnormal crud/corrosion layer isdefined by a formation that increases the calculated fuel average temperature by more than 25 'Cbeyond the design basis calculation...."AREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page 2-6Additional information was included in response to the RODEX4 SER restriction on addressing crud. Theinformation states "As our clad measurements [The RODEX4 SER restriction specifies that specific analyses would be required if an abnormal crud orcorrosion layer is observed at any given plant. The SER restriction defines an abnormal crud or corrosionlayer by a formation that increases the calculated fuel average temperature by more than 25 °C beyondthe design basis calculation. As part of the approved RODEX4 methodology, analyses are alreadyperformed for each cycle. Where plant specific measurements indicating abnormal crud are obtained,analyses for that plant are based on the plant specific data. If liftoff levels are found to be greater thanthose used in the RODEX4 corrosion model benchmark, then a plant-specific crud thickness will be inputto encompass the total liftoff thickness. The crud input will serve to satisfy the SER restriction on thedesign basis crud layer in cases where abnormal crud is encountered.Since the corrosion model was benchmarked to measurements that include some amounts of existingcrud, there is an inherent assumption [] The small temperature difference is otherwise not significant to fuel rod performance or safety.If abnormal levels of crud are encountered, the selection of the crud thermal conductivity input becomesmore important. The thermal conductivity of a crud layer depends on the composition, porosity, structure,and operating conditions and it continues to be a subject of development. Values of crud thermalAREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 2-7conductivity obtained from literature range from approximately 0.7 W/(m*K) up to nearly a factor of tengreater. Thermal conductivity values are more typically greater than 0.8 and less than that of oxide (i.e.,less than 2.0 W/(m*K)). Unless specific crud characterization and/or measurements are available for aplant, a value of [ ] will be used. Note that the combined layer of oxide and crud includes theselection of a conservative oxide thermal conductivity that contributes to the composite thermalresistance.In the transition to the ATRIUM 1OXM design at Monticello, []Crud is considered primarily a part of the plant operating environment that is mainly a function of the plantwater chemistry conditions.To assess the levels of crud at Monticello, Xcel provided AREVA with a prior water chemistry evaluationwhich contained the conclusion that current water chemistry conditions do not reduce fuel reliabilitymargins according to industry water chemistry guidelines.Visual examination data of fuel from past operating cycles in Monticello were provided to AREVA forreview. One examination was on two fuel assemblies that had operated for four cycles. Photos revealedcrud levels that visually appear consistent (i.e., no worse) in comparison to visuals at other plants withAREVA fuel that had previously been judged to have normal (i.e., low) crud levels as based on liftoffmeasurement data.During the re-channeling of eight fuel assemblies over the past year, Xcel characterized the appearanceof the fuel as having normal, uniform crud loading with no indications of abnormal crud layers.Based on the above information available from the Monticello plant, the crud conditions were taken to bewithin normal levels experienced by plants used for the RODEX4 corrosion model benchmarking.Therefore, no additional crud input was necessary to account for a design basis crud level.SNPB RAI-4: ANP-3092P, Section 2.0It has been stated in Section 2.0 that the ATRIUM IOXM fuel assemblies are hydraulically compatible withthe co-resident GE14 fuel design for the entire range of licensed power-to-flow operating map. Pleaseclarify this statement and elaborate whether the hydraulic compatibility between ATRIUM IOXM andGE14 fuel designs for all combinations of power-to-flow operating map for MELLLA + and extended flowwindow (EFW) during transition cycles as well as for full-core ATRIUM I OXM at MNGP.AREVA ResponseTables 3.7 and 3.8 of ANP-3092P provide thermal-hydraulic results for transition cycles at 100% corepower / 100% core flow (rated core flow at rated EPU power) and 59.2% core power / 43.3% core flow(minimum pump speed on MELLLA line), respectively. Table 1, Table 2, and Table 3 provide similarthermal-hydraulic results at 100% core power / 105% core flow (maximum core flow at rated EPU power),100% core power 1 80% core flow (minimum core flow at rated EPU power on MELLLA+ line) and 82.5%AREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 2-8core power / 57.4% core flow (minimum core flow at the MELLLA+ boundary). The results presented inTable 1, Table 2, and Table 3 demonstrate the thermal-hydraulic design criteria are satisfied for full-coreand transition core configurations. Differences in core average results (core pressure drop and corebypass flow) between ATRIUM 1OXM and GE14 results are within the range considered compatible.I] The Critical Power Ratio (CPR) results of the ATRIUM 1OXM andGE14 indicate ATRIUM 1OXM fuel will not cause thermal margin problems for the coresident fuel design.The results presented in ANP-3092P and the additional results provided in Table 1, Table 2, and Table 3demonstrate the hydraulic compatibility between ATRIUM 1OXM and GE14 fuel designs for the licensedpower/flow map for MNGP during transition cycles as well as for full-cores of ATRIUM 1OXM fuel.AREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page 2-9Table I Monticello Thermal-Hydraulic Results atRated Conditions (100%P / 105%F) forTransition to ATRIUM 10XM FuelIAREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page 2-10Table 2 Monticello Thermal-Hydraulic Results atRated Conditions (100%P / 80%F) forTransition to ATRIUM 1OXM FuelIAREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page 2-11Table 3 Monticello Thermal-Hydraulic Results atOff-Rated Conditions (82.5%P / 57.4%F) forTransition to ATRIUM 1OXM Fuel]AREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 2-12SNPB RAI-5: ANP-3092P, Section 3.1Please explain the details of test data reduction process and its modification to account for [[]].AREVA ResponseIIAREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page 2-13IAREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page 2-14ISNPB RAI-6: ANP-3092P. Sections 3.2 and 3.36.a) It appears that the NRC staff safety evaluation (SE) report for Revision 3 of EMF-2209(P)(A),"SPCB Critical Power Correlation" is a combination of the SE for EMF-2209(P) Revision 2Addendum 1. In the Revision I of EMF-2209 safety evaluation, the NRC staff has imposed fourconditions when the SPCB correlation is used for licensing applications.Describe how these conditions are satisfied when the SPCB correlation is used to compute thethermal margin performance for GE14 fuel at MNGP.AREVA ResponseThe four conditions and how they are satisfied when the SPCB correlation is applied to AREVA fuel areaddressed in Section 2-22 of ANP-3224P (which was provided with the fuel transition LAR, Reference 1Enclosure 6). The application of the SPCB correlation to Monticello GE14 fuel (SPCB/GE14) followstopical report EMF-2245(P)(A). The ranges of applicability identified in the first three conditions in EMF-2209(P)(A) represent the ranges of parameters available to AREVA when SPCB was developed forAREVA fuel. The ranges of some parameters for which GE14 critical power was available were reduced.Therefore, when applying SPCB to GE14 fuel, the ranges of applicability are equal to or more restrictivethan the ranges of applicability identified in the four conditions in EMF-2209(P)(A).Condition 1 states the SPCB correlation is applicable with a design local peaking factor no greater than1.5. SPCB/GE14 is applicable with a design local peaking factor no greater than 1.427.AREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 2-15Condition 2 states "If in the process of calculating the MCPR safety limit, the local peaking factor exceeds1.5, an additional uncertainty of 0.026 for ATRIUM-9B and 0.021 for ATRIUM-10 will be imposed on a rodby rod basis." For SPCB/GE14, an additional uncertainty of 0.068 is imposed when the local peakingfactor exceeds 1.427 for the MCPR safety limit.Condition 3 states "The SPCB correlation range of applicability is 571.4 to 1432.2 psia for pressure, 0.087to 1.5 Mlb/hr-ft2 for inlet mass velocity and 5.55 to 148.67 Btu/Ibm for inlet subcooling." Based on therange of parameters for GE14 critical power information that were available to AREVA, the ranges ofapplicability for SPCB/GE14 are reduced to 800 to 1300 psia for pressure, 0.18 to 1.5 Mlb/hr-ft2 for inletmass velocity and 5.55 to 100 Btu/Ibm for inlet subcooling. Appendix G of ANP-3224P presents aconservative method for extending the low pressure boundary for SPCB/GE14 to 571.4 psia.6.b) EMF-2245(P)(A) Revision 0, "Application of Siemens Power Corporation's Critical PowerCorrelations to Co-Resident Fuel" describes the processes for the application of approved SPC(AREVA) BWR critical power correlations to the co-resident fuel remaining from prior reloads(such as GE14 fuel at MNGP). The two processes that are presented in the topical report are theindirect process and the direct process.Please identify which process is used at MNGP and describe the steps used in the processimplemented.AREVA ResponseThe indirect approach was used. AREVA previously used the indirect approach to develop additiveconstants and uncertainties in order to apply the SPCB correlation for GE14 fuel. For Monticello, GNFprovided correspondence stating that the critical power performance previously furnished for GEl4 fuel isapplicable for the GE14 fuel with shorter heated length in Monticello without adjustment. Therefore, theSPCB correlation is being applied to the MNGP GE14 fuel with the same additive constants anduncertainties that were applied to GE14 fuel during the AREVA fuel transition for a previous BWR. Theonly difference is the conservative method for extending the low pressure boundary as described inAppendix G of ANP-3224P.Development of the SPCB/GE14 correlation followed the steps described in Section 3.1 of EMF-2245(P)(A). The following overview is provided.1. The SPCB correlation was used to predict critical power for each set of condition for which GE14critical power was known. The range of these conditions established the ranges of applicabilitymentioned in response to RAI-6.a.2. The calculated critical power data were then used to establish the appropriate additive constants forthe co-resident fuel using the approved procedures for the critical power correlation being used.3. The additive constant uncertainty (i.e., standard deviation) for the GE14 fuel was determined using theequations identified on page 3-2 of EMF-2245(P)(A). The additive constant uncertainty forSPCB/GE14 was calculated to be [ ] for theAREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 2-16MCPR safety limit. As mentioned in response to RAI-6a, [] for the MCPR safety limit.6.c) From Tables 3-5 and 3-6 of ANP-3092(P), it can be seen that for the transition core off-ratedconditions (59.2%P/43.3%F) thermal-hydraulic results critical power ratio (CPR) has a highermargin than the CPR for rated conditions.Please explain the reason for this higher margin for off-rated conditions of power and flow.AREVA ResponseSince the calculations in Tables 3-5 and 3-6 use the same assembly peaking factor, the assembly poweris Table 3-6 is 59.2% of the assembly power in Table 3-5. CPR is the ratio of the assembly power whichwould result in boiling transition at some location (Critical Power) to the actual assembly power. The CPRmargin increased in Table 3-6 because the reduction in critical power resulting from the reduction inassembly flow, along with the changes in pressure, subcooling etc, was less than 59.2%. In other words,the reduction in assembly power increased the margin to boiling transition more than the reduction inassembly flow and other factors decreased the margin to boiling transition.Similar results are seen for actual operating conditions of GE14 fuel in MNGP. One example is when thecore maneuvered from 100% pre-EPU core power and 88.3% core flow to 78% pre-EPU core power and66.6% core flow. At the initial conditions the CPR of the limiting GE14 assembly was 1.846 and at thenext steady state conditions the CPR of the same assembly was 2.080.SNPB RAI-7: ANP-3092P, Section 3.4Please provide details of the analysis to determine the impact of rod bow on thermal margin at lower andhigher exposures of ATRIUM IOXM fuel at MNGP.AREVA ResponseThe approach described below was used in the MNGP analysis for the impact of rod bow on thermalmargin over the entire range of exposures. AREVA uses the NRC approved correlation described intopical report XN-75-32(P)(A) Supplement 1 (Reference 7). The correlation was developed [] at the request of the NRC as discussed in Reference 7. [IAREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page 2-17]AREVA's BWR rod bow CPR penalty was derived using open literature data, Based on this data, it wasconcluded that thermal margins were not substantially reduced for closures as low as 0.03 inch.AREVA's model application for ATRIUM 10 type fuel was presented in an informational submittal to theNRC (Reference 11).The MCPR penalty (decrease in MCPR) versus rod bow (% closure) for the ATRIUM 10 fuel design ispresented in the Figure 2. To assure that this model is conservative, AREVA ran a CHF test on anATRIUM-10 bundle in which two rods were welded together. The measured MCPR penalty is also shownin Figure 2. This shows that the measured MCPR penalty for 100% closure was less than the MCPRpenalty predicted by the model.The impact of rod bow on thermal margin of ATRIUM 1OXM fuel at MNGP is provided in response toSNPB RAI-10.AREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page 2-18IFigure 2 -MCPR Penalty Model vs. Test DataSNPB RAI-8: ANP-3119P, Table 2-1Table 2-1 of ANP-3119P lists the active fuel length for full length rods and part length rods of A TRIUMIOXM design as 145.24 inches and 75.0 inches, respectively. The active fuel length for full length rods forA TRIUM 1OXM fuel design approved for an earlier plant application and for fuel rods for another plantcurrently under review is 150 inches.Please provide a summary of the impact on mechanical, thermal and neutronics characteristics of thereduced active full length designed for MNGP unit as compared to AREVA NP's previous designs, bothapproved and under review at the agency.AREVA ResponseThe 3% difference in full-length fuel rod length between 145.24 inches and 150 inches is considered to bea minor difference. There is not expected to be any significant impact on mechanical, thermal, andneutronic characteristic associated with this difference in full-length fuel rod length. Assemblies with145.24 inch full-length fuel rods and assemblies with 150 inch full-length rods, meet all requirements ofthe NRC approved Generic Mechanical Design Criteria (ANF-89-98(P)(A)) and are, therefore, fullyqualified for use in Boiling Water Reactors. All AREVA Analyses supporting the Monticello LAR areperformed explicitly modeling the full-length rods as 145.24 inches. There are no evaluations thatextrapolate performance based on analyses of 150 inch full-length rods.AREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page 2-19SNPB RAI-9: ANP-3119P, Sections 3.3.1 and 3.3.99.a) Please provide a summary of the stress evaluation analysis performed to confirm the designmargin and the establishment of a baseline that added accident loads. Also provide the results ofthe evaluation analysis that show that the assembly structural component criteria are maintainedunder normal and faulted conditions.AREVA ResponseAs discussed in ANP-3119P Section 3.3.1, [ ] the fuel assembly structuralcomponents do not receive significant loads during normal and AOO conditions. [] No analyses are performed to confirm design margin under normaloperating and AOO conditions [ ] The following text describes howAREVA's approved methodology was conservatively applied for the Monticello ATRIUM 1OXM stressevaluation.To ensure the structural integrity of [ " Section III of the ASME Boiler and PressureVessel code (Reference 13) is used to establish acceptable design limits. To evaluate the stresses undernormal operating conditions, [] The maximum normal operation [] for MNGP is then compared against the limit to ensurethat adequate margin is maintained.To evaluate the stress under AOO and accident conditions, [IAREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 2-20For the [ ] the normal operating stresses [] The design margin is confirmedby comparing the resulting stress to the design limit as defined by Section III of the ASME Boiler andPressure Vessel code (Reference 13).Information on the stress evaluation results and comparison to the load limits that show that the assemblystructural component criteria are maintained under faulted conditions can be found in Table 3-1,Section 3.4.4 of ANP-3119P.9. b) Please provide supporting analysis and results to show that structural integrity of the waterchannel is maintained under loads generated during normal operation and anticipated operationaloccurrences (AO0).AREVA ResponseThe answer to this RAI was included in the response to SNPB RAI-9 (9.a).SNPB RAI-10: ANP-3119P, Section 3.3.5Please provide the details of the analysis and results from the analysis that assure that the lateral creepbow of the fuel rods is not of sufficient magnitude to impact on thermal margins for the ATRIUM I OXM fueldesign at MNGP.AREVA ResponseThere is no CPR penalty due to lateral creep bow of the fuel rods before an assembly average exposureof 34.7 GWd/MTU. ATRIUM 1OXM fuel will not reach this exposure during their first cycle of irradiation.The rod bow CPR penalty reaches 0.01 at 42.8 GWd/MTU. By this assembly exposure in the equilibriumcycle, for all three of the ATRIUM 1OXM neutronic designs, the CPR margin of the cycle limitingATRIUM 1OXM assembly had increased to 0.40 or larger. This illustrates that rod bow does not affectthermal margins due to the lower powers achieved by high exposure assemblies (as mentioned onpage 4-7 of Reference 11).Additional information about rod bow is provided in SNPB RAI-7.SNPB RAI-11: ANP-3119P, Section 3.3.8Please provide typical calculations that show that there are large margins to assembly lift-off under normaloperating conditions and faulted conditions for the transition cycles at MNGP mixed core conditions.AREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page 2-21AREVA ResponseA liftoff calculation was performed for the ATRIUM 1OXM under the reactor conditions of MonticelloNuclear Generating Plant, to ensure that the following design criteria established in ANF-89-98(P)(A)(Reference 14) are met:-For normal operation and anticipated operational occurrences (AOO), the submerged fuel assemblyweight, including the channel must be greater that the hydraulic loads.-For accident (faulted) condition the normal hydraulic plus additional accident loads shall not cause theassembly to become disengaged from the fuel support, to assure that control blade insertion is notimpaired.The calculation [ITable 4 MNGP ATRIUM IOXM Fuel AssemblyIAREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page 2-22ISNPB RAI-12: ANP-3119, Table 3-2Please provide a description of the analysis and its results how oxidation and hydriding were accountedfor in the stress and fatigue analyses.AREVA ResponseCorrosion of structural components must be conservatively bounded in strength and fatigue calculationsto account for the material loss that occurs during oxidation. This is [] were used to show that the fuelchannel can withstand the duty cycle loads. There are no hydrogen uptake limits for structuralcomponents. However, AREVA performed calculations to ensure that the ductility of the water channelcould support a 1% strain limit (or higher) consistent with the requirements on fuel rod cladding.Calculations have estimated the hydrogen absorption in a water channel to be at 286 ppm at the end oflife. Measurements on unirradiated cladding have shown that 1% strain is achievable even with hydrogenlevels in excess of 500 ppm. A large margin is therefore maintained.SNPB RAI-13: ANP-3224P Section 2-10NRC staff approved Topical Report, EMF-93-177(P)(A) Revision I with three SER restrictions and withthree more restrictions carried over from Revision 0 of the TR.Please provide details of how the SER restrictions Numbers 1, 2, 4, 5 and 6 are met for theimplementation of EMF-93-177(P)(A) at MNGP.AREVA ResponseThe response below provides the details showing how the SER restrictions numbers 1, 2, 4, 5 and 6 aremet for the implementation of EMF-93-177(P)(A) at MNGP.AREVA Inc. Responses to RAI from SNPB onMNGP Transition to AREVA FuelANP-3289NPRevision 0Page 2-23Restriction number 1:"The fuel channel TR (Technical Report) methods and criteria may be applied to fuel channel designssimilar to the configuration of a square box with radiused corners open at the top and bottom ends.The wall thickness shall fall within the range of current designs. The channels shall be fabricated fromeither Zircaloy-2 or Zircaloy-4. AREVA will not use Zircaloy material for channels which has lessstrength than specified in the TR, and if the strength of material is greater than that in the TR, AREVAwill not take credit for the additional strength without staff review."The ATRIUM 10XM delivered to MNGP used the advanced fuel channel with 0.1 inch thick corners and0.075 inch thick side walls made from Zircaloy-4 sheets. These [IRestriction number 2:"Updates to channel bulge and bow data are permitted without review by the NRC staff; however,AREVA shall resubmit the channel bulge and bow data statistics if the two-sigma upper and lowerbounds change by more than one standard deviation."AREVA [the analyses.] The D-lattice plants approved models were used inRestriction number 4:"The allowable differential pressure loads and accident loads should bound those of the specificplant."AREVA performed [IResults were summarized in Tables 3-1, 3-2 and 3-3 of Reference 12.Restriction number 5:"Lattice dimensions should be compatible to those used in the analyses reported such that theminimum clearances with control blades continue to be acceptable."AREVA's analyses used the MNGP lattice dimensions.Restriction number 6:"Maximum equivalent exposure and residence time should not exceed the values used in theanalyses."AREVA's analyses []AREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 2-24SNPB RAI-14: ANP-3224P Section 2-11NRC staff approval of Topical Report BAW-10247PA was subject to five SER restrictions.Please provide details of how the SER restriction Number 5 regarding crud deposition is handled in thecase of fuel transition to A TIUM IOXM at MNGP.AREVA ResponsePlease see the response to SNPB RAI-3. The topic is similar so a consolidated response was provided.SNPB RAI-15: ANP-3224P Section 2-15Topical report EMF-2158(P)(A) was approved by the NRC staff subject to six SER restrictions of whichnumber 6 is "AREVA shall notify any customer who proposes to use the CASMO-4/MICROBURN-B2code system independent of any AREVA fuel contract that conditions I through 4 above must be met.AREVA's notification shall provide positive evidence to the NRC that each customer has been informedby AREVA of the applicable conditions for using the code system."Please explain how this SER restriction is implemented for the ATRIUM I OXM fuel transition process atMNGP.AREVA ResponseSER Restriction 6 of EMF-2158(P)(A) is applicable when the CASMO-4/MICROBURN-B2 code system issupplied independent of an AREVA fuel contract. In the case of the MNGP LAR theCASMO-4/MICROBURN-B2 code system is supplied integral to an AREVA fuel contract. Therefore, SERRestriction 6 does not apply.SNPB RAI-16: ANP-3224P, Sections 2-20, 2-21 and 2-22ANF-524(P)(A) was approved by the NRC staff subject to four SER restrictions. The SPCB correlationreplaced the ANFB correlation.16.a) Is the SPCB correlation applied to the co-resident GE14 fuel in the MNGP core during thetransition to ATRIUM I OXM?AREVA ResponseYes, please refer also to the response to SNPB RAI-6.16.b) Please explain how the SER restriction number 3 regarding the CPR channel bowing penalty fornon-ANF fuel (co-resident GE14 fuel) is applied for the MNGP core.AREVA ResponseThe fuel channel bow model in ANF-524(P)(A) was based on assembly exposure. SER Restriction 3 is"The CPR penalty bowing penalty for non-ANF fuel should be made using conservative estimates of thesensitivity of local power peaking to channel bow". The sensitivity of local peaking to channel bow wasAREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 2-25calculated with a 4 bundle CASMO "colorset" calculation. Monticello MCPR Safety Limit calculations usethe methodology described in Reference 16 (this was mentioned at the end of Section 2-20 inANP-3224P).This methodology implements a model which calculates channel bow based on thedifference in fluence between opposite sides of the fuel channel. The sensitivity of local peaking tochannel bow is calculated with MICROBURN-B2 3D nodal powers for the bowed condition. This changein the modeling of channel bow is summarized in Reference 16 (Figure 2-1 in ANP-10307Q1P, AREVAMCPR Safety Limit Methodology Responses to RAIs).AREVA requested and GNF provided information about channel bow for the GE14 fuel design in MNGP.This was compared to the channel bow predicted by applying the AREVA fluence gradient model to theGNF fuel assemblies. Based on this comparison, [] These adjustments weredetermined and applied so that the predicted channel bow for the GNF fuel using the AREVA bow modelwas either in alignment with, or conservative relative to the GE channel bow information for all exposures.SNPB RAI-17: ANP-3224P. Section 2-2517.a) One of the modifications made with regards to ANP-10307PA Revision 0 is to address a concernwith the application of the fuel channel bow standard deviation when the fluence gradient iscomputed to exceed the bound of the channel measurement database. As a consequence, thefuel channel bow standard deviation component of the channel bow model uncertainty used byANP-10307PA to determine the Safety Limit Minimum Critical Power Ratio was increased by theratio of channel fluence gradient to the channel fluence gradient bound of the channelmeasurement database, when applied to channels with fluence gradients outside the bounds ofthe measurement database from which the model uncertainty was determined.Please explain the details of the above modification that was implemented in ANP-10307PA inconnection with the fuel transition at MNGP unit.AREVA ResponseThe maximum fluence gradient for each assembly in Cycle 28 was calculated by MICROBURN-B2. A fewassemblies were predicted to experience a fluence gradient slightly larger than the fluence gradient usedin the model benchmark. The approach previously developed to address this NRC concern (refer toRAI-17.b) was applied to MNGP. Specifically,AREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 2-26]17.b) During review of Brunswick Steam Electric Plant (BSEP) ATRIUM IOXM fuel transition LAR, theNRC staff determined that the predictive model for channel bow was validated against anempirical data that was not bounding of BSEP's expected performance. To resolve this issue, thelicensee for BSEP agreed to increase the channel bow uncertainty in the SLMCPR calculation forthe most severely deflected fuel channels. In view of the excessive channel bow that occurred atBSEP, a license condition was proposed for BSEP Units I and 2 in connection with the use ofAREVA channel bow model outside the range of the channel bow measurement database fromwhich its uncertainty was quantified (

Reference:

Letter, BSEP 13-0002, from Michael J. Annacone(Duke Energy) to NRC, "Supplement to License Amendment Request for Addition of AnalyticalMethodology Topical Report to Technical Specification 5.6.5, CORE OPERATING LIMITSREPORT (COLR), and Revision to Technical Specification 2.1.1.2 Minimum Critical Power RatioSafety Limit," Duke Energy, January 22, 2013).Confirm whether a similar license condition is required for the MNGP unit for the fuel transition.AREVA ResponseBy virtue of its inclusion in the analysis submitted in the LAR (i.e., Section 2-25 of ANP-3224), thelicensee has accepted the channel bow uncertainty in the SLMCPR calculation as an element of themethodology. Therefore, no license condition is required for MNGP.SNPB RAI-18: ANP-3224P, Section 2-29Please provide a latest revision of the methodology and code manual for XCOBRA- T code that iscurrently used in the thermal hydraulic core analysis for the fuel transition in the MNGP core.AREVA ResponseThe latest revision of the XCOBRA-T methodology is identified in Section 2-29 of ANP-3224P(Reference 17). During a conference call with the NRC reviewer on January 27, 2014 regarding this RAI,the NRC reviewer determined that the XCOBRA-T methodology did not need to be provided.The latest revision of the XCOBRA-T code manual (Reference 18) is being provided with these RAIresponses. This is being provided for information only and is proprietary in its entirety.SNPB RAI-19: ANP-3224P, Section 2-15 and Appendix AIt appears from the above mentioned sections of your submittal forATRIUM 1OXM fuel transition atMNGP unit that the licensee has used Topical Report, EMF-2158(P)(A) to calculate radial and axial powerdistribution measurement uncertainties.Please provide details of analyses, calculations, and the database information used to establish theseuncertainties. Also, please confirm that the uncertainties (including those for TIP distribution uncertainties)calculated for MNGP unit is in line with the uncertainties that are listed in Chapter 9 of EMP-2158(P) (A).AREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 2-27AREVA ResponseRadial and axial power distribution measurement uncertainties used in the MNGP safety limit analysesare not taken from EMF-2158(P)(A). These uncertainties are based upon the GARDEL core monitoringuncertainties. The GARDEL core monitoring power distribution measurement uncertainties were providedin Reference 19. In Enclosure 3 of Reference 21, Xcel Energy discussed its approach to evaluate theimpact of explicitly accounting for the 25% grace period for the LPRM calibration interval.SNPB RAI-20: ANP-3224P, Appendix AIt is stated in Section A l of the Appendix A that the methods used in CASMO-4 are state of the art. Themethods used in MICROBURN-B2 are state of the art. The methodology accurately models a wide rangeof thermal hydraulic conditions including EPU and extended power/flow operating map conditions.Please explain how the CASMO-4 and MICBURN-B2 calculations are applied to the extended power/flowoperating map conditions.AREVA ResponseCASMO-4 calculations are performed for various void fractions and fuel temperatures. This data is usedby MICROBURN-B2 to construct cross sections consistent with the operating state-point conditions ofpower, flow and pressure. The specific values of power, flow and pressure are input values used in thecalculation. The core design engineer utilizes values appropriate for the state-point to be analyzed withinthe constraints of the licensing restrictions of the power/flow map. Mutual solutions of the powerdistribution and flow distribution are used to determine the conditions of each node in the core.SNPB RAI-21: ANP-3224P, Appendix DIn page D-2 it is stated that the multi-rod database used in the [[fl. As a result, the multi-rod database and prediction uncertaintiesare not available to AREVA. However, the correlation has been independently validated by AREVAagainst public domain multi-rod data and proprietary data collected for prototypical ATRIUM-IC andATRIUM IOXM test assemblies. Selected results for the ATRIUM-IC test assembly are reported in thepublic domain in Reference 42.The NRC staff would like to review the data that has validated the correlation and requests a copy of theReference 42 that is listed in ANP-3224P.AREVA ResponseThe requested reference is provided as an enclosure to the Xcel Energy response.SNPB RAI-22: ANP-3138PNRC staff has performed the review of the supplemental topical report, ANP-10298PA Revision 0,Supplement IP Revision 0, "Improved K-Factor Model forACE/ATRIUM 10 XM Critical PowerCorrelation," December 2011. If the final approval is done before the implementation of the fuel transitionat MNGP unit, it may be advantageous to list the approved supplement to the MNGP COLR/TS.AREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 2-28Note: This RAI is advanced information for the licensee to monitor development of the approval processfor the supplemental TR.AREVA ResponseWhen NRC approval of ANP-10298PA Revision 0, Supplement 1P Revision 0, "Improved K-Factor Modelfor ACE/ATRIUM 10 XM Critical Power Correlation," December 2011 is obtained, Xcel Energy will providerevised documentation to incorporate this document in the license basis for the Fuel Transition LAR.AREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 3-13.0 References1. License Amendment Request for Transition to AREVA ATRIUM 10XM Fuel and AREVA SafetyAnalysis Methodology, July 15, 2013, MNGP L-MT-13-055, ML1 3200A1 85.2. Monticello Nuclear Generating Plant -Request for Additional Information Regarding LicenseAmendment Request to Transition to AREVA ATRIUM 1OXM Fuel and Safety Analysis Products(TAC No. MF2479), ML13200A185.3. BAW-10247PA Revision 0, "Realistic Thermal-Mechanical Fuel Rod Methodology for BoilingWater Reactors," AREVA NP Inc., February 2008.4. XN-NF-79-59(P)(A), Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies,Exxon Nuclear Company, November 1983.5. XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors,THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company,January 1987.6. ANP-3224P Revision 2, Applicability of AREVA NP BWR Methods to Monticello, AREVA NP,June 2013.7. XN-75-32(P)(A) Supplements 1 through 4, Computational Procedure for Evaluating Fuel RodBowing, Exxon Nuclear Company, October 1983. (Base document not approved.)8. XN-NF-82-06(P)(A) Supplement 1 Revision 2, Qualification of Exxon Nuclear Fuel for ExtendedBurnup, Supplement 1, "Extended Burnup Qualification of ENC 9x9 BWR Fuel", May 1988.9. EMF-85-74(P), Revision 0, Supplement I(P)(A) and Supplement 2(P)(A), RODEX2A (BWR) FuelRod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.10. "Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thermal MarginCalculations for Light Water Reactors (Revision 1)," NRC Report dated February 16, 1977.11. EMF-95-52(P) Revision 1, Fuel Design Evaluation for Siemens Power Corporation A TRIUMTM-1OBWR Reload Fuel, April 1998, transmitted to the NRC by Siemens Power Corporation Letter,"Design Evaluations for SPC ATRIUMTM-9B and ATRIUMTM-10 Fuel", April 8, 1998, (NRC:98:021).12. ANP-3119P Revision 0, "Mechanical Design Report for Monticello ATRIUM IOXM FuelAssemblies", AREVA Inc., October 2012.13. ASME Boiler and Pressure Vessel Code, Section III, Division 1, American Society of MechanicalEngineers.14. ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR FuelDesigns, Advanced Nuclear Fuels Corporation, May 1995.15. EMF-93-177(P)(A), Revision 1, Mechanical Design for BWR Fuel Channels, August 2005.16. ANP-10307PA Revision 0, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors,"AREVA NP, June 2011.AREVA Inc. ANP-3289NPResponses to RAI from SNPB on Revision 0MNGP Transition to AREVA Fuel Page 3-217. XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2 Revision 0, XCOBRA-T: AComputer Code for BWR Transient Thermal-Hydraulic Code Analysis, Exxon Nuclear Company,February 1987.18. EMF-CC-167(P) Revision 8, XCOBRA-T Theory, Programmer's and Users Manual, AREVA NP,November 2011.19. Response to Requests for Additional Information (RAI) for the License Amendment Request toRevise the Minimum Critical Power Ratio Safety Limit in Reactor Core Safety Limit 2.1.1.2(TAC No. ME4790), February 8, 2011, MNGP L-MT-11-009, ML110450240.20. EMF-95-52(P) Revision 1, Fuel Design Evaluation for Siemens Power Corporation A TRIUMTM-1OBWR Reload Fuel, April 1998, (transmitted to the NRC by Siemens Power Corporation Letter,"Design Evaluations for SPC ATRIUMTM-9B and ATRIUMTM-10 Fuel", April 8, 1998, NRC:98:021).21. AREVA ATRIUM 1 OXM Fuel Transition -Responses to Request for Additional Information(TAC MF2479), January 31, 2014, MNGP L-MT-14-003, ML14035A297.AREVA Inc. }}