ML14027A705
| ML14027A705 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 12/17/2013 |
| From: | Luminant Generation Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML14028A551 | List: |
| References | |
| CAW-13-3866, CP-201301411, LAR 13-01, NF-TB-13-119, TAC MF1365, TAC MF1366, TXX- 13182, TXX-13182 | |
| Download: ML14027A705 (37) | |
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Attachment 2 to TXX- 13182Page I of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)1. For the limiting Group F2 fuel assembly selection, the only design that does not include axial blanketswas not included. The licensee stated that "the combination of this fuel type's [I1I it to be bounded by I[ ]l." Please list the parameters thatcaused this assembly to not be considered.Response:]aC The differences between the unblanketed fuel assembly and the Group F2design basis assembly are outlined in Table 1-1.Table 1-1: Comparison of the Unblanketed and Design Basis Fuel Assembly DesignsParameter Unblanketed Design Group F2 Design BasisBlanket Enrichment' (wt% 235U) 1.6, 2.4, and 3.1 ac[ ]95/95 Fuel Density (%TD) I ]I [. ]cacBurnable Absorber Loading 20 WABA fingers [ ]cReactor Power (MWth) 3458 3612Notes:1. The blanket enrichment refers to the enrichment of the top and bottom 6" of both the blanketedand unblanketed assembly.Unblanketed fuel assemblies were only made in enrichments of 1.6, 2.4, and 3.1 weight percent (wt%) 235U.]'c The parameters used forthe unblanketed fuel depletion calculations are outlined in Table 1-2.Table 1-2: Depletion Parameters for Unblanketed Assembly DesignParameter Unblanketed Design Unblanketed Depletion ParameterReactor Power (MWth) 3458 [ ac95/95 Fuel Density (%TD) [ [ ]acAssembly Temperature Profile' [ ]ax [ ]8Burnable Absorber Loading 20 WABA fingers ac[ ]Notes:1. [a8c
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2 to TXX- 13182Page 2 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)The spent fuel isotopics generated in the depletion calculations were imported to KENO V.a and the reactivity ofthe unblanketed fuel assemblies were calculated. The reactivity of the unblanketed fuel assembly was comparedto the reactivity of the F2 design basis assembly. The results of the reactivity comparison for burnup bin 1 areprovided in Table 1-3.LTable 1-3: Reactivity Comparison at 3.0 wt% 23SU for Burnup Bin IJ acThe results in Table 1-3 show that the Group F2 design basis assembly is more reactive than the unblanketed fuelassembly in burnup bin 1.]axCReferences:1. WCAP- I 7728-P, Revision 1, "Comanche Peak Nuclear Power Plant Units I and 2 Spent Fuel PoolCriticality Safety Analysis," October 2013.
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2 to TXX-13182Page 3 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)2. In the selection of the limiting Group F2 assembly, the licensee explained that the combination of WetAnnular Burnable Absorber (WABA) and Integral Fuel Burnable Absorber (IFBA) conservatively boundsthe other designs that only use one or the other (i.e., WABA or IFBA). Please provide the results of theanalysis that demonstrates this is true.Response:Westinghouse has provided below a reactivity comparison of the design basis burnable absorber loading to theburnable absorber loadings covered in Table 6-16 of Reference 1. [I as
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2 to TXX- 13182Page 4 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)Past Use of the 64 IFBA/24 WABA Burnable Absorber LoadingI a~c__Table 2-1: Reactivity Comparison of the Design Basis and 64 IFBAI24 WABA Past Use AssembliesI a,cI ac
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2 to TXX-13182Page 5 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)a,cjjTable 2-2: Reactivity Comparison of the Design Basis and 64 IFBA/24 WABA Future Use Assemblies IL---m
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2 to TXX- 13182Page 6 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)]aa,c-[Table 2-3: Reactivity Comparison of the Design Basis and 156 IFBA Rod Assemblies at 3.0 wt% 235UI-
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2 to TXX- 13182Page 7 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)Table 2-4: Reactivity Comparison of the Design Basis and 156 IFBA Rod Assemblies at 4.0 wt% 235U Aa,ca,c_1 Table 2-5: Reactivity Comparison of the Design Basis and 156 IFBA Rod Assemblies at 5.0 wt% 235UIIa~c
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2 to TXX- 13182Page 8 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)References:1. WCAP-17728-P, Revision 1, "Comanche Peak Nuclear Power Plant Units 1 and 2 Spent Fuel PoolCriticality Safety Analysis," October 2013.
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2 to TXX-13182Page 9 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)3. WCAP-1 7728-P, "Comanche Peak Nuclear Power Plant Units I and 2 Spent Fuel Pool Criticality SafetyAnalysis" (proprietary, not publicly available) (Enclosure 2 to letter dated March 28, 2013), Section 4.2.1,states that II II. Pleaseexplain if this is considered to be conservative because HI JI.Response:References:1. WCAP- I 7728-P, Revision 1, "Comanche Peak Nuclear Power Plant Units I and 2 Spent Fuel PoolCriticality Safety Analysis," October 2013.
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2 to TXX- 13182Page 10 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)4. WCAP-17728-P, Section 4.2.3.2, discusses axial moderator temperature profile selection, but does notdiscuss how PARAGON treats the moderator density. Please explain if this is the bounding moderatordensity profile used in the same manner as the bounding moderator temperature profile.Response:The axial moderator density is calculated by the FIGHTH code, as part of the core design package used atWestinghouse, which solves the steady-state heat equation, given the values of linear heat rate, burnup, flow rate,and moderator temperature.a~cReferences:1. WCAP-1772-P, Revision 1, "Comanche Peak Nuclear Power Plant Units 1 and 2 Spent Fuel PoolCriticality Safety Analysis," October 2013.
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2 to TXX- 13182Page II of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)5. To compensate for a lack of critical experiments containing fission products, ItII as described in WCAP-] 7728-P, Section5.3.2.1.4, and is assessed based on preliminary research performed by Oak Ridge National Laboratory(ORNL). More recent research performed by ORNL in NUREG/CR-7109, "An Approach for ValidatingActinide and Fission Product Burnup Credit Criticality Safety Analyses -Criticality (krr) Predictions,"April 2012 (ADAMS Accession No. ML121 16A128), shows that 1.5 percent of the minor actinide andfission product worth (treated as a bias) is acceptable to account for the lack of a sufficient number ofapplicable critical experiments containing minor actinides and fission products. Applying theNUREG/CR-7109 recommendations for determining uncertainty attributed to fission product and minoractinide validation gaps, the NRC staff estimates that the licensees approach would produce a non-conservative uncertainty estimate by approximately 100 to 200 percent millirho (pcm), but the actualvalue could be larger. II11 please provide a justification for not including the minor actinides andfor not applying the results of the more recent research in the manner recommended.Response:I] a,c
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2 to TXX-13182Page 12 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)a,cTable 5-1: 1]~I a,c
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2 to TXX- 13182Page 13 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)p aca,c-11 Table 5-2: [lacI_III a~c
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2 to TXX-13182Page 14 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)References:1. WCAP-17728-P, Revision 1, "Comanche Peak Nuclear Power Plant Units I and 2 Spent Fuel PoolCriticality Safety Analysis," October 2013.2. ORNL/TM-12973, "Sensitivity and Parametric Evaluations of Significant Aspects of Burnup Credit forPWR Spent Fuel Packages," Oak Ridge National Laboratory, July 1996.
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2 to TXX- 13182Page 15 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)6. For the minimum margin case in the accident analysis, which occurs with a multiple assembly misload,there is approximately 100 pcm to the regulatory keff limit of 0.95. Since the margin for the limitingaccident case is minimal, please confirm that there is no increase in the total bias and uncertainty term dueto not considering the presence of soluble boron when determining the combined bias and uncertaintyterm for this minimum margin case. The NRC staff notes that WCAP- I 7483-P, "WestinghouseMethodology for Spent Fuel Pool Rack Criticality Analysis," December 2011 (proprietary, not publiclyavailable), which WCAP-17728-P is based on, recommends using a 500 pcm bias to account for anypotential increase.Response:To confirm that the presence of soluble boron has not increased the total bias and uncertainty term, the highestworth biases and uncertainties were recalculated in the limiting multiple misload accident model. Thesecalculations were performed with a 2400 ppm soluble boron concentration and a fresh 5.0 wt% assemblymisloaded in []axI Table 6-1: Recalculated Bias and Uncertainty Term for Borated Accident ConditionI ac
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2 to TXX-13182Page 16 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)IaxReferences:1. WCAP- I 7728-P, Revision 1, "Comanche Peak Nuclear Power Plant Units I and 2 Spent Fuel PoolCriticality Safety Analysis," October 2013.2. WCAP-17483, "Westinghouse Methodology for Spent Fuel Pool and New Fuel Rack Criticality SafetyAnalysis," December 2011.
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2 to TXX-13182Page 17 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)7. A publication titled, "Atomic Weights of the Elements: Review 2000," from the Journal of Pure andApplied Chemistry, Volume 75, Number 6, pp. 683-800, from 2003 shows that the B-10 isotopic fractioncan be as low as 0.192 in general for naturally occurring terrestrial samples, with one study showingsamples with a B-10 isotopic fraction as low as 0.1893. Since the margin for the limiting accident case isminimal, pleasejustify the B-10 isotopic fraction of I1 11.Response: 'The boric acid used in the Reactor Coolant System and Spent Fuel Pools at CPNPP is purchased in accordancewith quality assurance standards, which require that the boric acid be "undepleted in boron 10 isotope", and thatthe manufacturer is required to supply an isotopic analysis report of the B-10 composition.A review of these records at CPNPP from 1992 to 2012 indicates that the boric acid utilized at CPNPP has beennear the nominal value of a 0.199 atom fraction. The average value of these receipts was 0.199 atom fractionB-10, and the minimum value in this 20 year period was 0.1973.To verify the current B-I 0 content in the Spent Fuel Pools, CPNPP performed B-10 isotopic analysis on watersampled from each pool on 10/31/2013. The isotopic analysis results demonstrate a B- 10 content of 0.1982 atomfraction for Spent Fuel Pool 1, and 0.1977 atom fraction for Spent Fuel Pool 2. These results demonstrate that theassumptions utilized in the analysis are conservative based on current conditions.Potential decreases in SFP B- 10 concentration may occur due to three possible sources:A. First, it is possible that Depleted Boron from the RCS may mix with the SFP water, lowering the B-10concentration. During a normal cycle of operation, the RCS Boron concentration is diluted to very lowvalues (<100 ppm). Prior to opening the transfer canal gates to allow core offload (therefore connectingthe RCS to the SFP), the RCS boron is increased to a value greater than 2400 ppm. This addition of freshboron typically establishes a B-I 0 concentration very near the natural value prior to mixing the RCS withthe SFP water. However, it is possible that a mid-cycle shutdown may result in a boration from a muchhigher initial concentration of depleted B-10, which may result in a lower than expected B-10concentration. This potential is addressed in the response to RAI #15, which demonstrates that when thetransfer system is opened, it is not feasible that the refueling cavity water could have a B-10 concentrationless than [ ]aCTo ensure the impacts of any abnormal RCS depletion scenarios on the SFP are well understood in thefuture, CPNPP will review the calculated B-10 concentration in the RCS each refueling outage (afterborating to >2400 ppm, but not including the fill of the Refueling Cavity). If the calculated value isbelow [ ]fc, a B-10 measurement will be performed on the Spent Fuel Pool afteradequate mixing time has occurred, but prior to the next refueling outage, to ensure the B- 10 value in theSFP has not significantly changed.B. Secondly, it is possible that the addition of fresh boric acid to the SFP, which was created from naturallyoccurring boron with abnormally low B-10 content, may reduce the SFP B-10 concentration. Unlike theRCS, there is rarely (if ever) a need to actively reduce or increase the SFP Boron concentration; therefore,fresh boric acid is rarely added to the Spent Fuel Pool. The Boron concentration cannot be reduced below2400 ppm due to restrictions of Technical Specification 3.7.16 and other administrative controls. Anypotential future addition of boric acid to the SFP (even at extremely low values of B-10 concentration)could only have a minor impact on B-10. For example, assume the SFP boron concentration is at 2400ppm, with a B-10 atom fraction of 0.197. If the SFP boron is significantly increased by 200 ppm usingfresh boric acid with an abnormally low atom fraction of 0.189, the resulting SFP B-10 content would bereduced to 0.1964 atom fraction. The SFP would need to be diluted and borated between 2400 ppm and
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2 to TXX- 13182Page 18 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)2600 ppm a total of 6 times, using fresh boric acid at 0.189 atom fraction for all boration, to reduce theSFP B-10 concentration below [ ]" It is concluded that it is not feasible for boricacid additions to invalidate the B-10 assumption utilized in the analysis.To ensure the impacts of any abnormal RCS boration scenarios on the SFP are well understood in thefuture, CPNPP will review the SFP Boron Measurement history each refueling outage. If the SFP boronvalues have experienced any increase of more than 100 ppm, a review of B-10 values for Boric Acidpurchased at CPNPP will be performed. If this review demonstrates that boric acid has been purchasedwhich has a B-J0 atom fraction below [ ]as, a B-10 measurement will be performed on the SpentFuel Pool prior to the next refueling outage.C. Lastly, depletion of the B-J0 due to neutron activity within the SFP may reduce the B-10 concentration.This potential is addressed in the response to RAI #15, and it is concluded that it is not feasible fordepletion in the SFP to invalidate the B-10 assumption utilized in this analysis. Therefore, no actions arenecessary to address this B-10 reduction potential.As discussed further in the response to Question #6, although the reactivity margin demonstrated in the limitingmisload analysis is minimal, the limiting case was performed using extremely conservative conditions which weredemonstrated to not be credible fuel accident conditions.
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2 to TXX-13182Page 19 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)8. Please provide clarification for the following items related to fuel handling:a. The last paragraph in WCAP-17728-P, Section 5.5.5, says that the inspection cells can only evercontain one fuel assembly at a time. However in WCAP-17728-P, Section 5.5.2, it states that upto two assemblies can be placed together in the sipping equipment. This appears to be conflictinginformation. Please provide clarification.b. If two assemblies are allowed in the inspection cells, please explain the physical means thatensure that at least one assembly pitch is always maintained between assemblies.c. Please explain the physical means that ensure that one assembly pitch is always maintainedbetween the inspection cell and the storage racks.d. There is a requirement for Region II that no fuel be placed in the interfacing row of the inspectioncell. Please explain why the same requirement does not exist for Region I (CPNPP, Units I and2, SFPs interface with Region 1).e. Please explain the meaning of the last sentence in Section 5.5.2, which states, "Note that it is alsoacceptable to perform these tasks with the section of the assembly that is being manipulatedabove the storage racks."Response to #8.a:The fuel sipping equipment discussed in this section is not placed inside the inspection cells, and the conditionsdescribed in Section 5.5.2 are not related to the inspection cells.The inspection cells are oversized SFP rack locations, which are designed to allow a fuel assembly to be loweredand rotated during fuel inspection activities, so that the full range of the assembly may be inspected withoutlowering and raising the underwater cameras. There are two inspection cell locations in SFPI, and a singleinspection cell location in SFP2, as shown in Figures 3-1 and 3-2 of WCAP-17728-P Rev. 1. Fuel assemblyInspection activities may be performed outside of the storage racks as described in 5.5.2. Even when using theInspection Cells to perform fuel inspections, the camera equipment would remain above the storage racks. If theassembly is partially lowered into the Inspection Cell during this activity (to aid in viewing the top sections of thefuel assembly), then the conditions and limitations for utilizing the Inspection Cells described in Section 5.5.5would apply.The Inspection Cell locations do NOT contain adequate room to place fuel sipping equipment. Section 5.5.2 ofWCAP-17728-P Rev. I states that during fuel sipping conditions, "the fuel assemblies are separated by at leastone assembly pitch via equipment design." This equipment design prevents the potential to insert this equipmentinto the inspection cells, since these cells are only 2 storage cells wide. Due to physical limitations of the racks,the type of equipment described in Section 5.5.2 could only be placed external to the storage racks, such as in theWet Cask Pit or above empty storage cell locations.Response to #8.b:Two assemblies are not allowed into the inspection cells. Use of the inspection cells is described in more detail inthe response to RAI 8.a.Response to #8.c:The requirement for utilizing the inspection cells is independent of the "one assembly pitch" requirementdescribed in Section 5.5.2. The inspection cell usage is discussed in Section 5.5.5, and the restrictions aredescribed in Section 6.3. The requirement to maintain empty locations around the Region II inspection cells alsoexisted with the previous Criticality Safety Analysis at CPNPP, and procedural requirements currently exist to
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2 to TXX- 13182Page 20 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)ensure storage cells adjacent to the inspection location are empty prior to use. The empty cells next to theinspection cell provides the physical barrier to assure "one assembly pitch" is maintained.Response to #8.d:For Region II, the restriction for inspection cell usage is described in 6.3, which states "fuel inspection cells inRegion 11 can only be used if no fuel is stored in the adjacent fuel storage cells." Similar to the requirements for"empty cells" in WCAP-17728-P Rev. I Figure 5-1, "Allowable Storage Arrays", this restriction applies toadjacent Region 11 fuel storage cells, and does not apply to the Interface of Region I. With the adjacent storagecells vacant, the Inspection Cell reactivity is bound by Storage Array II-E, which allows for I out of 4 storage,surrounded by empty cells.For the Region I Inspection Cell, there are no restrictions described in 6.3 since the reactivity of any fuel assemblyplaced into this storage cell is bound by normal fuel storage in Region 1. Note that the Region I inspection cell isNOT simply a void in the racks where 4 cells were removed, but is an oversized cell location, including walls(with gaps to the adjacent cells) and BORAL neutron absorber panels. Therefore, there is no need to restrictadjacent storage locations in Region I, and no need to impose any special interface restrictions for the Region I /Region II interface.Response to #8.e:The preceding discussion describes how the reactivity of cleaning, inspection, reconstitution, and sippingactivities are bound by the analysis for Array II-E. Since Array II-E applies for fuel storage inside the Region IIracks, and the Region II racks do not contain neutron poison material, then it can be concluded that this analysis isbounding for a fuel assembly outside of the storage racks, when this assembly is more than I assembly pitch awayfrom any other fuel assemblies.This statement is a simple acknowledgment that although the Array II-E analysis is used to bound the activitiesdescribed in this section, cleaning and fuel inspection activities are nonnally performed above the plane of thestorage racks, or in an area which does not contain storage racks, such as the fuel transfer canal. In some cases,fuel is inserted into a storage cell, and the inspection or other activities will take place on the upper part of the fuelassembly as the fuel is raised or lowered. These activities are bound by the Array II-E storage analysis.
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2 to TXX- 13182Page 21 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)9. For Array II-A, depicted in WCAP-17728-P, Section 5.2, II]I It was not obvious to the NRC staff that the unconsidered misload scenario for ArrayI1-A would be non-limiting for the misload analysis. Since this misload scenario is credible, pleasedemonstrate that a fresh assembly misload in Array 11-A is non-limiting.Response:] axCase 16 F 4 46 6 X 4Case 36 6 4 4F 6 X 4Case 26 6 4 46 F X 4Case 4F 6 4 46 6 X 4Figure 9-1 Misload Cases in Array II-AII Table 9-1: 1I 8cIIII II aII-I aE,c
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2 to TXX- 13182Page 22 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)II Table 9-2: 1rzI c Ia,,References:1. WCAP- I 7728-P, Revision 1, "Comanche Peak Nuclear Power Plant Units I and 2 Spent Fuel PoolCriticality Safety Analysis," October 2013.
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2 to TXX-13182Page 23 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)10. Please explain if the misplacement of a fuel assembly is only possible in what is shown in WCAP-17728-P, Figures 3-1 and 3-2, as the inspection cell regions. If other regions, other than the inspectioncell area, are open to the misplacement of fuel assemblies, please identify them.Response:It is not physically possible to place a fuel assembly between the outer boundaries of the storage racks and theSFP walls, or to misplace a fuel assembly within the outer boundaries of the storage racks in locations other than astorage cell or the inspection cells.The CPNPP Region I cells in each Spent Fuel Pool contain a small number of 'cut-off cells near the Spent FuelPool Swing Gate (these locations are identified in Figures 3-1 and 3-2 of WCAP-l7728-P Rev. 1). These cells arephysically identical to a normal storage cell, including placement and length of the BORAL neutron absorbers,with the exception that the top 12" of the cell (which is above the active fuel region and BORAL poison region)has been removed to prevent contact with the SFP Swing Gate as it opens into the pool. Although it is physicallypossible to insert fuel into these locations, fuel assemblies are restricted from these cells due to limitations of theSwing Gate.
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2 to TXX- 13182Page 24 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)11. There appears to be a typo in Section 5.7.4, which states, "... provides 0.05 Ak of reactivity suppression."It appears that this value should be corrected to 0.005 Ak.Response:The statement "...provides 0.05 Ak of reactivity suppression." in Reference I is correct as listed. The value0.05 Ak refers to the approximate margin from criticality provided by 320 ppm of soluble boron. Table 5-20 ofReference I shows the neutron multiplication factor (k~ff) of each array assuming a soluble boron concentration of320 ppm in the spent fuel pool, including biases and uncertainties but not including administrative margin. Thedifference between a keff of 0.995 (1.0 -0.005 Ak administrative margin) and the keff values shown in Table 5-20gives the reactivity worth of 320 ppm of soluble boron for each configuration. For Array 11-A this value is 0.995 -0.94474 = 0.05026 Ak.References:1. WCAP- I 7728-P, Revision 1, "Comanche Peak Nuclear Power Plant Units I and 2 Spent Fuel PoolCriticality Safety Analysis," October 2013.
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2 to TXX- 13182Page 25 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)12. In Region 1, the rack modules are designed with a flux trap, but WCAP-I7728-P does not discuss creditof the flux trap gap during a seismic event due to structural considerations. Since it is possible for theflux trap gap size to change during seismic activity, please explain why full credit of the Region I rackmodule flux trap gap during a seismic event is appropriate.Response:The CPNPP storage racks are designed to meet the seismic Category I requirements of RG 1.13 and RG 1.29."Water Gap Flats" are part of the rack design, and are welded between the storage cells to ensure the spacing ismaintained at all times, including during a seismic event.As stated in the "No Significant Hazards Consideration" section of LAR 13-01, "the margin of safety with respectto mechanical, material or structural considerations is not changed by this proposed License AmendmentRequest." This structural design of the racks has been considered previously, and approved by the NRC, inAmendment 87 of the CPNPP Operating License (reference ML012560143). In the "Mechanical, Material andStructural" section of Attachment 2 to the related License Amendment Request, the following statements aremade:"The Region I / Region I1 racks have a sufficient margin of safety against tilting and deflection or movementduring a seismic event. The Region I / Region II racks do not impact each other or the pool walls, damage fuelassemblies, or cause criticality concerns during a postulated seismic event.""The Region I / Region 1I rack weld stresses at the connections (e.g., baseplate-to-rack, baseplate-to-pedestal, andcell-to-cell connections) were calculated under the dynamic loading conditions. All of the calculated weld stressesare less than the corresponding allowable stresses specified in the ASME Code, indicating that the weldconnection design of the rack is adequate."Therefore, previously approved analysis demonstrates that the flux trap gap assumed in WCAP-17728-P Rev. 1will remain intact during a seismic event, and crediting this gap in the criticality analysis is appropriate.Also, note that due to the fact that the SFP Storage Racks are not restrained or physically attached to the SFPliner, it is possible the inter-rack gaps may be reduced during a seismic event. Section 5.7.4 of WCAP-17728-PRev. I addresses the potential for the rack modules to slide together during a seismic event and the potentialreduction in the space between the modules.
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2 to TXX- 13182Page 26 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)13. Please provide details on how B-10 areal density manufacturing variation, absorber thickness variation,and degradation of BORAL absorption ability over time is accounted for in the criticality analyses. As aminimum, provide answers to the following clarification questions.a. Is the II II areal density the adjusted neutron absorber loading accounting only for theabsorber thickness tolerance (i.e. I[ II)?b. Please explain why the neutron absorber thickness tolerance is not listed in Tables 3-8 or 5-3 eventhough it was accounted for by adjusting the neutron absorber areal density. What is this tolerancebased on?c. Please explain if the B4C density bias is applied in Tables 5-8 and 5-13 based on a toleranceperturbation for B-10 areal densities of I[ I] and [1 i] as suggested by Table5-3, Note 1.d. The values in Table 5-3 imply B-10 areal density values, but Tables 5-8 and 5-13 list a B4C densitybias. Please explain if the B4C density values were adjusted in the KENO models to match the targetB-10 areal densities given in Table 5-3. Please provide the KENO material specifications used tomodel Boral.e. Is II Ithe minimum certified B-10 areal density?f. Section 5.1.2.4, "Impact of Potential BORAL Blistering," paragraph 3 states that the areal density isadjusted from 1I 11 to II II to account for IIJj -this adds additional confusion as Table 5-3 does notmention [i 11 Please explain how is this adjustment accounted for inthe criticality safety analysis?Response:a. A tolerance on the Boral thickness was not specified by the manufacturer.b. Please see response to a.c. [d.I as
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2 to TXX- 13182Page 27 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)I]a Ce. Yes, []" is the minimum certified areal density.f. Please see response to a.
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2 to TXX-13182Page 28 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)14. In WCAP-17728-P, Section 5.3.2.1.2, it is stated that the burnup measurement uncertainty is taken to be"the reactivity change associated with a 11 11 change in burnup," however it is not explained why avalue of 11 1] is appropriate for this uncertainty term. Please provide justification for use of a II IIchange in burnup for the burnup measurement uncertainty.Response:Multiple industry studies have been performed to detenrmine the accuracy of reactor burnup records.NUREG/CR-6998 (Reference 1) involved evaluation of several thousand in-core measured assembly burnupvalues. Reference 1 states that "utility records for fuel burnup are accurate for individual spent fuel assemblies toat least 5% of "true" assembly bumup." EPRI Report TR- 112054 (Reference 2) presents an additional studybased on in-core measurement comparisons, which confirms that the burnup measurement uncertainty is less than1 ] C. According to Section 7.2 of Reference 1 and Section 4.3 of Reference 2, the uncertainty in the utility-assigned bumup measurement values is less than the [ ]"C value used in Reference 4 when computer models,correctly normalized to start-of-cycle conditions and adjusted periodically on the basis of in-core measurements,are used. Additionally, Westinghouse performed a study of assembly power uncertainty which is documented inReference 3, confirming that deviation of the measured values from the reactor records is within [ PC.Because Comanche Peak has used core-follow systems such as CONFORM and the BEACONTM CoreMonitoring System throughout its operating history to adjust the estimated assembly burnups based on flux mapresults, the use of a burnup measurement uncertainty value of [ ]IG for the discharge burnups is conservative.References:1. NUREG/CR-6998, "Review of Information for Spent Nuclear Fuel Burnup Confirmation," December2009.2. TR-1 12054, "Determination of the Accuracy of Utility Spent-Fuel Burnup Records," EPRI, July 1999.3. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988.4. WCAP- I 7728-P, Revision 1, "Comanche Peak Nuclear Power Plant Units 1 and 2 Spent Fuel PoolCriticality Safety Analysis," October 2013.
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2 to TXX-13182Page 29 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)15. The revised Bases for TS 3.7.16 states that "the effect of B-10 depletion on the boron concentration formaintaining keff less than or equal to 0.95 is accounted for in 1[ ]]," however, B-10depletion is not discussed in WCAP- 17728-P. Please explain how is it accounted for.Response:]8CThe other potential source of depleted 10B in the SFP is from mixing SFP and RCS water. The soluble '°B in theRCS depletes during operation due to the absorption of neutrons. As the 10B depletes, the soluble boronconcentration decreases as reactor operators remove soluble boron from the RCS to maintain criticality. The 10Bat% is lowest (most depleted) at the end of cycle when the plant shuts down for refueling, this is also the time thatthe RCS contains the least amount of soluble boron. Once the plant has shut down and is in Mode 5 or 6 forrefueling, the RCS is borated to at least the minimum allowable SFP soluble boron concentration (2400 ppm).The RCS boration is performed using undepleted soluble boron from the Boron Accumulator Tank (BAT).I acReferences:1. WCAP-17728-P, Revision 1, "Comanche Peak Nuclear Power Plant Units I and 2 Spent Fuel PoolCriticality Safety Analysis," October 2013.
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2 to TXX- 13182Page 30 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)16. If the [1 1] is included in the area of applicability (AOA) analysis, pleaseexplain why Boral minimum areal density is not included.Response:a~c
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2 to TXX- 13182Page 31 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)17. WCAP-17728-P, Section 6. 1, contains tables with various coefficients to be used with an equationrelating the initial fuel enrichment to the minimum burnup for fuel assembly loading into the variousstorage arrays (i.e. these tables define, by curve fit, the various burnup and enrichment loading curves).The methodology for curve fitting is not explained. It is not clear if the curves are designed to passdirectly through the explicit bumup and enrichment points or if they are somehow bounded. Pleaseprovide additional details on how the burnup and enrichment equations are developed based on the aboveconsiderations.Response:The following criteria were considered when generating the curve fits and fitting coefficients found in ReferenceI, in order to ensure conservatism, while maintaining simplicity.I a~cReferences:1. WCAP- I 7728-P, Revision I, "Comanche Peak Nuclear Power Plant Units I and 2 Spent Fuel PoolCriticality Safety Analysis," October 2013.
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2 to TXX- 13182Page 32 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)18. WCAP-17728-P, Section 6.2.1, states that outlier assemblies can be stored in arrays that do not requireburnup credit without performing an analysis, but must be below the maximum fresh fuel enrichment.This general allowance is potentially problematic. For example, if a future fuel design incorporated higherenrichment blankets, this would allow more reactive fuel to be stored in both Region I and II without anyre-analysis.This issue is also described in letter dated July 16, 2013 (supplemental information submitted by thelicensee in response to NRC staff letter dated June 27, 2013, Item 4 (ADAMS Accession No.MLI3175A225)), indicating that a new analysis would have to be performed before fuel can be loaded inthe SFP if the fuel assembly in question cannot be categorized as Group F1 or F2. The response statedthat only the first 5 parameters of WCAP- I 7728-P, Table 6-16 should be evaluated, however parameters6 and 7 -[I I -should also be evaluated as they arealso based on the fuel design II I1 Please explain whyparameters 6 and 7 have been excluded.Response:IIIaxC
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2 to TXX- 13182Page 33 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)I Table 18-1: 1Ia.ca,ca~cReferences:1. WCAP-17728-P, Revision 1, "Comanche Peak Nuclear Power Plant Units 1 and 2 Spent Fuel PoolCriticality Safety Analysis," October 2013.
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2 to TXX-13182Page 34 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)19. It is evident that significant rodded operation is not anticipated by virtue of the imposed 0.1 GWd!MTUmax rodded operation limit in the AOA analysis; additionally it is stated that rods were not inserted intothe core more than 20 cm at any given time. Consequently, past and current fuel cycles appear to bereasonably covered by the as-defined AOA; however, this has not been demonstrated for future fuelcycles that do not fall within the AOA.WCAP- 1 7728-P, Section 6.2.1, regarding future rodded operation not covered by the CPNPP AOA, statesthat IIJ] However, in the supplemental information provided by letter dated July 16,2013 (Item 4), the licensee stated the following with respect to depletion parameters of future fuelassemblies that fall outside of the defined AOA:If the parameter only impacts the fuel depletion assumptions of WCAP-17728-P and the fuel needs to bestored in the SFP, it shall be placed in either Region I or in Array II-E in Region II.This statement is not consistent with the methodology presented in WCAP-17728-P, Section 6.2.1mentioned above and specifically imposes a requirement for a burned fuel assembly outside of the AOAbased on depletion characteristics to be stored in either Region I or in Region 1I (i.e., the 1 out of 4storage array configuration) as if it were fresh. Based on the conflicting information above, please provideclarification for how outlier fuel assemblies that do not meet the fuel depletion criteria in the AOAdefined by Table 6-16 in WCAP- 17728-P will be stored for CPNPP.Response:CPNPP approach for treating HFP rodded operation assumption outliers:The CPNPP response to Question #4 in the supplemental information was accurate at the time, since CPNPPoriginally planned to treat all 'outliers', or fuel assemblies which do not satisfy the depletion assumptions ofWCAP-17728-P Rev. I Table 6-16, the same. This included outliers which did not satisfy the assumption forMaximum HFP Rodded Operation. CPNPP was not, at the time, planning on utilizing the provision in theanalysis which states "assemblies which are classified as outlier assemblies because of HFP Rodded Operationcan be stored using burnup credit if the bumup accrued during rodded operation is not credited."The decision to not utilize this provision was made based on CPNPP's understanding at that time that outliers tothe AOA (Area of Applicability) would be evaluated per 50.59, and evaluations for outliers would not requireNRC approval (assuming the supporting analysis demonstrated that the resulting reactivity impacts were notadverse). Based on clarifying discussions both internally and with the NRC, CPNPP now understands the AOA isconsidered part of the supporting methodology; therefore the required evaluation of AOA outliers will requireNRC approval of the supporting analysis.Based on this revised understanding, CPNPP plans on utilizing the provision in WCAP-1 7728-P Rev. I forapplying burnup credit to store fuel assemblies which are classified as outlier assemblies solely due to HFPRodded Operation. For these assemblies, burnup which is accrued during HFP rodded conditions will NOT becredited in the Technical Specification Surveillance, but all other burnup accrued during the cycle will becredited.For all other depletion parameters in Table 6-16 of WCAP-17728-P Rev.1 (Including HFP Rodded Operationparameter in combination with another depletion parameter), CPNPP will not apply any credit for burnup in theTechnical Specification Surveillance, unless prior approval is obtained from the NRC.
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2 to TXX-13182Page 35 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)Justification for not crediting burnup accrued during HFP rodded operation:As implied in Section 6.2.1 of WCAP-17728-P Rev. 1, the depletion which occurs under rodded conditions maybe ignored to ensure that the reactivity assumed in the Technical Specification surveillance is bounded by theanalysis. Section 5.8 of WCAP-17728-P Rev. 1 describes how the reactivity of an assembly experiencing roddedoperation can increase relative to an assembly which does not experience rodded operation, due to several factors.Due to this potential, the depletion which occurs under these conditions cannot be credited in the surveillancewithout further analysis. This comparison between rodded and unrodded operation describes the relativereactivity between two fuel assemblies with the same burnup value, and does not imply that the fuel assemblyreactivity, at any axial location, could actually increase due to depletion under rodded conditions. In other words,rodded depletion, even under the most extreme rodded conditions, cannot increase the reactivity at any axiallocation in the fuel assembly. Therefore, it is inherently conservative to ignore this depletion, and only credit theburnup which accrued under normal operating conditions.For example, assume that a Fuel Assembly is depleted for 200 days at normal HFP ARO conditions. Thisassembly then experiences 100 days of HFP rodded conditions. The TS 3.7.17 surveillance for the fuel assemblywould only credit 200 EFPD (Effective Full Power Days) of burnup, since the 100 days of depletion duringrodded operation is not credited. Due to the additional uncredited depletion time, the assembly is inherently lessreactive than it was following the 200 days of normal HFP ARO operation; specifically it will be less reactive ateach axial location (including areas which were covered by control rods, which experienced some amount of fueldepletion even under these abnormal conditions), regardless of the final axial burnup profile.In a more realistic scenario, the fuel will likely experience additional unrodded depletion following the period ofrodded operation. Section 5.8 of WCAP-17728-P Rev. 1 describes that "once the RCCA has been withdrawnfrom the assembly, power preferentially moves to the under-depleted top of the assembly and over time the axialburnup profile developed will return to a profile typical of unrodded operation." If 100 days of rodded operationfor the fuel assembly in the example above had occurred prior to, or during, the 200 days of HFP ARO operation,the axial burnup profile would eventually return to a profile similar to a fuel assembly depleted under unroddedconditions, due to the preferential depletion of the under-depleted axial locations. In this case, the reactivityassumed by crediting only 200 EFPD of bumup in the surveillance bounds the actual reactivity at all axiallocations.Note that the NRC has reviewed and approved a similar treatment of Rodded Operation in a recent SafetyEvaluation related to SFP criticality analysis. The proposed treatment of rodded operation burnup is similar to thecommitments documented in "PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS I AND 2 -ISSUANCE OF AMENDMENTS RE: SPENT FUEL POOL CRITICALITY CHANGES (TAC NOS. ME6984AND ME6985)," dated August 29, 2013, Reference ML13241A383. On page 9 of the supporting SafetyEvaluation Report (Enclosure 3 to this document), the following Prairie Island commitment is described:"In conjunction with implementation of the amendment, procedures will be revised to require anassessment ofa fiuel assembly's exposure to rodded power operation in the core prior to moving that fielassembly into the spent fiel pool (SFP) storage racks. If an assembly experiences more than 100megawatt day per metric ton uranium (MWd/MTU) of core average full-power rodded operationexposure, this exposure experienced while rodded will not be credited for determining the coefficientsused to categorize fiel assemblies as described in WCAP-1 7400-P. "
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2 to TXX- 13182Page 36 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)In the discussion following the commitment, the Safety Evaluation concluded that the approach was acceptable:"The current version of the commitment does not allow for significant amounts of rodded operation to beconsidered when determining whether afuel assembly meets the storage requirements. Therefore, theNRC concludes that this commitment acceptably accounts for the variability of rodded operation."Treatment of HFP rodded operation outliers in past cycles:Section 5.8 of WCAP-17728-P Rev.] discusses the potential impacts of Rodded Operation on fuel reactivity, andincludes a statement that "Comanche Peak has not operated at full power with control rods inserted a significantlength... Therefore, there is no significant bumup accrued during depletion with RCCAs inserted in the activefuel height, and no need to account for these effects in burnup limits contained within this analysis."Section 6.2 summarizes the Area of Applicability of the analysis, which includes key assumptions utilized in theanalysis which will be "confirmed for each cycle of operation to assure that the results presented here remainvalid." The Maximum HFP Rodded Operation in the AOA is 0.1 GWD/MTU/cycle.CPNPP has recently completed a more detailed review of past plant history to confirm the AOA assumptionsrelated to Maximum HFP Rodded Operation. Although the vast majority of past CPNPP operation wasperformed under unrodded conditions, and the general statements made in Section 5.8 regarding past operationare valid, there are several past cycles which do not satisfy the conservative threshold of 0.1 GWD/MTU/cycle offull power operation below 210 steps (from Table 6.16).Because the depletion analysis did not specifically address the time spent during HFP Rodded Operation, CPNPPplans on treating rodded fuel assemblies from past cycles in a manner identical to future cycles, i.e., the fueldepletion which occurred during HFP Rodded Operation will NOT be credited in the Technical SpecificationSurveillance for any cycle which accrued more than 0.1 GWD/MTU of HFP Rodded Operation.The administrative controls and Configuration Confirmation Software tools described in Enclosure I of LAR 13-01 will incorporate limitations to ensure that the appropriate burnup is credited for fuel assemblies which haveexperienced HFP Rodded Operation beyond the low threshold required by the AOA. As described in Enclosure Iof LAR 13-01, only use of QA-controlled software is permitted for performance of SR 3.7.17.1.
Attaclm~ent 2 to TXX- 13182Page 37 of 37Comanche Peak Responses to LAR 13-01 Request for Additional Information (RAI) -Round 2(Non-Proprietary)20. Please explain why the axial burnup profile evaluated in the AOA is not defined by Table 6-16.Response:[aReference:1. WCAP-17728-P, Revision 1, "Comanche Peak Nuclear Power Plant Units I and 2 Spent Fuel PoolCriticality Safety Analysis," October 2013.